Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases, 56819-56864 [2015-23193]

Download as PDF Vol. 80 Friday, No. 181 September 18, 2015 Part V Nuclear Regulatory Commission mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR Part 50 Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases; Proposed Rule VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 PO 00000 Frm 00001 Fmt 4717 Sfmt 4717 E:\FR\FM\18SEP5.SGM 18SEP5 56820 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC–2011–0088] RIN 3150–AI97 Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases Nuclear Regulatory Commission. ACTION: Proposed rule. AGENCY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to incorporate by reference seven recent editions and addenda to the American Society of Mechanical Engineers (ASME) codes for nuclear power plants and a standard for quality assurance. The NRC is also proposing to incorporate by reference four ASME code cases. This action is in accordance with the NRC’s policy to periodically update the regulations to incorporate by reference new editions and addenda of the ASME codes and is intended to maintain the safety of nuclear power plants and to make NRC activities more effective and efficient. DATES: Submit comments by December 2, 2015. Comments received after this date will be considered if it is practical to do so, but the NRC is able to ensure consideration only for comments received on or before this date. ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject): • Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC–2011–0088. Address questions about NRC dockets to Carol Gallagher; telephone: 301–415–3463; email: Carol.Gallagher@nrc.gov. For technical questions contact the individuals listed in the FOR FURTHER INFORMATION CONTACT section of this document. • Email comments to: Rulemaking.Comments@nrc.gov. If you do not receive an automatic email reply confirming receipt, then contact us at 301–415–1677. • Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 301– 415–1101. • Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, ATTN: Rulemakings and Adjudications Staff. • Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 20852, between 7:30 a.m. and 4:15 p.m. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 SUMMARY: VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 (Eastern Time) Federal workdays; telephone: 301–415–1677. For additional direction on obtaining information and submitting comments, see ‘‘Obtaining Information and Submitting Comments’’ in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: Daniel I. Doyle, Office of Nuclear Reactor Regulation, telephone: 301– 415–3748, email: Daniel.Doyle@nrc.gov; or Keith Hoffman, Office of Nuclear Reactor Regulation, telephone: 301– 415–1294, email: Keith.Hoffman@ nrc.gov. Both are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001. SUPPLEMENTARY INFORMATION: Executive Summary A. Need for the Regulatory Action The NRC is proposing to amend its regulations to incorporate by reference seven recent editions and addenda to the ASME codes for nuclear power plants and an ASME standard for quality assurance. The NRC is also proposing to incorporate by reference four ASME code cases. This proposed rule is the latest in a series of rulemakings to amend the NRC’s regulations to incorporate by reference revised and updated ASME codes for nuclear power plants. The ASME periodically revises and updates its codes for nuclear power plants by issuing new editions and addenda, and this rulemaking is in accordance with the NRC’s policy to update the regulations to incorporate by reference those new editions and addenda. The incorporation by reference of the new editions and addenda will maintain the safety of nuclear power plants, make NRC activities more effective and efficient, and allow nuclear power plant licensees and applicants to take advantage of the latest ASME codes. The ASME is a voluntary consensus standards organization, and the ASME codes are voluntary consensus standards. The NRC’s use of the ASME codes is consistent with applicable requirements of the National Technology Transfer and Advancement Act. Additional discussion of voluntary consensus standards and the NRC’s compliance with the National Technology Transfer and Advancement Act (NTTAA) is set forth in Section VIII of this notice, ‘‘Voluntary Consensus Standards.’’ B. Major Provisions Major provisions of the proposed rule include: PO 00000 Frm 00002 Fmt 4701 Sfmt 4702 • Incorporation by reference of ASME codes into NRC regulations and delineation of NRC requirements for the use of these codes (including conditions). • Incorporation by reference of various versions of quality assurance standard NQA–1 into NRC regulations and approval for their use. • Incorporation by reference and approval of four ASME Code Cases. C. Costs and Benefits The NRC prepared a draft regulatory analysis to determine the expected costs and benefits of the proposed rule. The regulatory analysis identified costs and benefits in a quantitative fashion as well as in a qualitative fashion. The analysis concluded that the proposed rule would result in net quantitative costs to the industry and the NRC. The proposed rule, relative to the regulatory baseline, would result in a net cost for industry of between $5.1 million based on a 7 percent net present value and $4.3 million based on a 3 percent net present value. The estimated incremental industry cost per reactor unit ranges from $49,000 based on a 7 percent net present value to $41,000 based on a 3 percent net present value. The NRC benefits from the proposed rulemaking alternative because of the averted cost of not reviewing and approving Code alternative requests on a plant-specific basis under § 50.55a(z) of title 10 of the Code of Federal Regulations (10 CFR). The NRC net benefit ranges from $1.4 million based on a 7 percent net present value to $1.9 million based on a 3 percent net present value. Qualitative factors which were considered include regulatory stability and predictability, regulatory efficiency, and consistency with the NTTAA Act of 1995, as amended. Table 44 in the draft regulatory analysis includes a discussion of the costs and benefits that were considered qualitatively. If the results of the regulatory analysis were based solely on quantified costs and benefits, then the regulatory analysis would show that the rulemaking is not justified because the total quantified benefits of the proposed regulatory action do not equal or exceed the costs of the proposed action. However, if the qualitative benefits (including the safety benefit, cost savings, and other nonquantified benefits) are considered together with the quantified benefits, then the benefits outweigh the identified quantitative and qualitative impacts. With respect to regulatory stability and predictability, the NRC has had a decades-long practice of approving and/ E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules or mandating the use of certain parts of editions and addenda of these ASME Codes in 10 CFR 50.55a through the rulemaking process of ‘‘incorporation by reference.’’ Retaining the practice of approving and/or mandating the ASME Codes continues the regulatory stability and predictability provided by the current practice. Retaining the practice also assures consistency across the industry, and provides assurance to the industry and the public that the NRC will continue to support the use of the most updated and technically sound techniques developed by the ASME to provide adequate protection to the public. In this regard, these ASME Codes are voluntary consensus standards developed by participants with broad and varied interests and have already undergone extensive external review before being reviewed by the NRC. Finally, the NRC’s use of the ASME Codes is consistent with the NTTAA, which directs Federal agencies to adopt voluntary consensus standards instead of developing ‘‘governmentunique’’ (i.e., Federal agency-developed) standards, unless inconsistent with applicable law or otherwise impractical. For more information, please see the draft regulatory analysis (Accession No. ML14170B104 in the NRC’s Agencywide Documents Access and Management System). mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Table of Contents I. Obtaining Information and Submitting Comments A. Obtaining Information B. Submitting Comments II. Background III. Discussion A. ASME BPV Code, Section III B. ASME BPV Code, Section XI C. ASME OM Code D. ASME Code Cases IV. Section-by-Section Analysis V. Generic Aging Lessons Learned Report VI. Specific Request for Comments VII. Plain Writing VIII. Voluntary Consensus Standards IX. Incorporation by Reference—Reasonable Availability to Interested Parties X. Environmental Assessment and Final Finding of No Significant Environmental Impact XI. Paperwork Reduction Act Statement XII. Regulatory Analysis: Availability XIII. Backfitting and Issue Finality XIV. Regulatory Flexibility Certification XV. Availability of Documents I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC–2011– 0088 when contacting the NRC about the availability of information for this proposed rule. You may obtain VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 information related to this proposed rule by any of the following methods: • Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC–2011–0088. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may obtain publiclyavailable documents online in the ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/ adams.html. To begin the search, select ‘‘ADAMS Public Documents’’ and then select ‘‘Begin Web-based ADAMS Search.’’ For problems with ADAMS, please contact the NRC’s Public Document Room (PDR) reference staff at 1–800–397–4209, 301–415–4737, or by email to pdr.resource@nrc.gov. For the convenience of the reader, instructions about obtaining materials referenced in this document are provided in the ‘‘Availability of Documents’’ section. • NRC’s PDR: You may examine and purchase copies of public documents at the NRC’s PDR, Room O1–F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. B. Submitting Comments Please include Docket ID NRC–2011– 0088 in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at https:// www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Background The ASME develops and publishes the ASME Boiler and Pressure Vessel Code (BPV Code), which contains requirements for the design, construction, and inservice inspection (ISI) of nuclear power plant components; and the ASME OM Code,1 1 The editions and addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants have had different titles from 2005 to 2012 PO 00000 Frm 00003 Fmt 4701 Sfmt 4702 56821 which contains requirements for inservice testing (IST) of nuclear power plant components. Until 2012, the ASME issued new editions of the ASME BPV Code every 3 years and addenda to the editions annually, except in years when a new edition was issued. Similarly, the ASME periodically published new editions and addenda of the ASME OM Code. Starting in 2012, the ASME decided to issue editions of its BPV and OM Codes (no addenda) every 2 years with the BPV Code to be issued on the odd years (e.g., 2013, 2015, etc.) and the OM Code to be issued on the even years (e.g., 2012, 2014, etc.). The new editions and addenda typically revise provisions of the Codes to broaden their applicability, add specific elements to current provisions, delete specific provisions, and/or clarify them to narrow the applicability of the provision. The revisions to the editions and addenda of the Codes do not significantly change Code philosophy or approach. It has been the NRC’s practice to establish requirements for the design, construction, operation, ISI (examination), and IST of nuclear power plants by approving the use of editions and addenda of the ASME BPV and OM Codes (ASME Codes) in § 50.55a. The NRC approves and/or mandates the use of certain parts of editions and addenda of these ASME Codes in § 50.55a through the rulemaking process of ‘‘incorporation by reference.’’ Upon incorporation by reference of the ASME Codes into § 50.55a, the provisions of the ASME Codes are legally-binding NRC requirements as delineated in § 50.55a, and subject to the conditions on certain specific ASME Codes’ provisions that are set forth in § 50.55a. The editions and addenda of the ASME BPV and OM Codes were last incorporated by reference into the regulations in a final rule dated June 21, 2011 (76 FR 36232), subject to NRC conditions. The ASME Codes are consensus standards developed by participants with broad and varied interests (including the NRC and licensees of nuclear power plants). The ASME’s adoption of new editions of, and addenda to, the ASME Codes does not mean that there is unanimity on every provision in the ASME Codes. There may be disagreement among the technical experts, including NRC representatives on the ASME Code committees and subcommittees, regarding the acceptability or desirability of a particular Code and are referred to collectively in this rule as the ‘‘OM Code.’’ E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56822 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules provision included in an ASMEapproved code edition or addenda. If the NRC believes that there is a significant technical or regulatory concern with a provision in an ASMEapproved Code edition or addenda being considered for incorporation by reference, then the NRC conditions the use of that provision when it incorporates by reference that ASME Code edition or addenda. In some cases, the condition increases the level of safety afforded by the ASME code provision, or addresses a regulatory issue not considered by the ASME. In other instances, where research data or experience has shown that certain Code provisions are unnecessarily conservative, the condition may provide that the Code provision need not be complied with in some or all respects. The NRC’s conditions are included in § 50.55a, typically in paragraph (b) of that regulation. In a Staff Requirements Memorandum (SRM) dated September 10, 1999, the Commission indicated that NRC rulemakings adopting (incorporating by reference) a voluntary consensus standard must identify and justify each part of the standard that is not adopted. For this rulemaking, the provisions of the 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section III, Division 1; and the 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section XI, Division 1, of the ASME BPV Code; and the 2009 Edition, 2011 Addenda, and 2012 Edition of the ASME OM Code that the NRC is not adopting, or partially adopting, are identified in the Discussion, Regulatory Analysis, and Backfitting and Issue Finality sections of this notice. The provisions of those specific editions and addenda and Code Cases that are the subject of this rulemaking that the NRC finds to be conditionally acceptable, together with the applicable conditions, are also identified in the Discussion, Regulatory Analysis, and Backfitting and Issue Finality sections of this notice. The ASME Codes are voluntary consensus standards, and the NRC’s incorporation by reference of these Codes is consistent with applicable requirements of the NTTAA. Additional discussion on NRC’s compliance with the NTTAA is set forth in Section VIII of this notice, ‘‘Voluntary Consensus Standards.’’ This proposed rule contains changes from a November 5, 2014, NRC final rule amending § 50.55a to, among other things, re-designate paragraphs within § 50.55a (79 FR 65776). The redesignation of paragraphs was needed to address the Office of the Federal Register’s requirements in 10 CFR part VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 51 applicable to incorporation by reference. For additional information on the November 2014 final rule, please consult the statement of considerations (preamble) for that final rule. III. Discussion The NRC regulations incorporate by reference ASME codes for nuclear power plants. The ASME periodically revises and updates its codes for nuclear power plants. This proposed rule is the latest in a series of rulemakings to amend the NRC’s regulations to incorporate by reference revised and updated ASME codes for nuclear power plants. This rulemaking is intended to maintain the safety of nuclear power plants and make NRC activities more effective and efficient. The NRC follows a three-step process to determine acceptability of new provisions in new editions and addenda to the Codes and the need for conditions on the uses of these Codes. This process was employed in the review of the Codes that are the subjects of this rule. First, the NRC staff actively participates with other ASME committee members with full involvement in discussions and technical debates in the development of new and revised Codes. This includes a technical justification of each new or revised Code. Second, the NRC committee representatives discuss the Codes and technical justifications with other cognizant NRC staff to ensure an adequate technical review. Third, the NRC position on each Code is reviewed and approved by NRC management as part of the rule amending § 50.55a to incorporate by reference new editions and addenda of the ASME Codes and conditions on their use. This regulatory process, when considered together with the ASME’s own process for developing and approving the ASME Codes, provides reasonable assurance that the NRC approves for use only those new and revised Code edition and addenda, with conditions as necessary, that provide reasonable assurance of adequate protection to public health and safety, and that do not have significant adverse impacts on the environment. The NRC reviewed changes to the Codes in the editions and addenda of the Codes identified in this rulemaking. The NRC concluded, in accordance with the process for review of changes to the Codes, that each of the editions and addenda of the Codes, and the 2008 Edition and the 2009–1a Addenda of NQA–1, are technically adequate, consistent with current NRC regulations, and approved for use with the specified conditions. The NRC proposes to amend its regulations to incorporate by reference: PO 00000 Frm 00004 Fmt 4701 Sfmt 4702 • The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition to the ASME BPV Code, Section III, Division 1 and Section XI, Division 1, with conditions on their use. • The 2009 Edition, the 2011 Addenda, and the 2012 Edition to Division 1 of the ASME OM Code, with conditions on their use. • ASME Standard NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications,’’ including several editions and addenda to NQA– 1 from previous years with slightly varying titles as identified in proposed rule language § 50.55a(a)(1)(v). More specifically, the NRC proposes to incorporate by reference the 1983 Edition through the 1994 Edition, the 2008 Edition, and the 2009–1a Addenda to the 2008 Edition of ASME NQA–1, with conditions on their use. • ASME BPV Code Case N–729–4, ‘‘Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI, Division 1,’’ ASME approval date: June 22, 2012, with conditions on its use. • ASME BPV Code Case N–770–2, ‘‘Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1,’’ ASME approval date: June 9, 2011, with conditions on its use. • ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ ASME approval date: October 16, 2012. • ASME OM Code Case OMN–20, ‘‘Inservice Test Frequency.’’ The current regulations in § 50.55a(a)(1)(ii) incorporate by reference ASME BPV Code, Section XI, 1970 Edition through the 1976 Winter Addenda; and the 1977 Edition (Division 1) through the 2008 Addenda (Division 1), subject to the conditions identified in current § 50.55a(b)(2)(i) through (b)(2)(xxix). The proposed amendment would revise § 50.55a(a)(1)(ii) to incorporate by reference the 2009 Addenda (Division 1) through the 2013 Edition (Division 1) of the ASME BPV Code, Section XI. It would also clarify the wording and add, remove, or revise some of the conditions as explained in this notice. The NRC proposes to revise § 50.55a(a)(1)(iv) to incorporate by reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division 1 of the ASME OM Code. Based on this revision, the NRC regulations would E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules incorporate by reference in § 50.55a the 1995 Edition through the 2012 Edition of the ASME OM Code. Each of the proposed NRC conditions and the reasons for each proposed condition are discussed below. The discussions are organized under the applicable ASME Code and Section. Please note that there is not a separate heading for ASME quality assurance standard NQA–1 because there are three separate discussions of NQA–1—one under the heading for ASME BPV Code, Section III, one under the heading for ASME BPV Code, Section XI, and one under the heading for ASME OM Code—because there are three proposed conditions related to NQA–1, one in each of those areas (paragraph (b)(1)(iv) for Section III, paragraph (b)(2)(x) for Section XI, and paragraph (b)(3)(i) for the OM Code). A. ASME BPV Code, Section III 10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section III The NRC proposes to clarify that Section III Nonmandatory Appendices are not incorporated by reference. This language was originally added in a final rule published on June 21, 2011 (76 FR 36232); however, it was omitted from the final rule published on November 5, 2014 (79 FR 65776). The NRC is correcting the omission by inserting ‘‘(excluding Non-mandatory Appendices)’’ in 10 CFR 50.55a(a)(1)(i). mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions The NRC proposes to identify prohibited subparagraphs and footnotes for each BPV Code edition and addenda in tabular form as opposed to the textual listing of the current regulation. No substantive change to the requirements is intended by this revision. The NRC believes that presenting the information in tabular form will increase the clarity and understandability of the regulation. Currently, § 50.55a(b)(1)(ii) includes a condition prohibiting the use of Footnote 11 from the 1989 Addenda through the 2003 Addenda or Footnote 13 from the 2004 Edition through the 2008 Addenda to Figures NC–3673.2(b)– 1 and ND–3673.2(b)–1 for welds with leg sizes less than 1.09 tn when using the ASME BPV Code, Section III, Division 1. These Code provisions provide stress indices for welded joints used in the design of Class 2 and Class 3 piping. The use of these indices is prohibited for welds with leg sizes less than 1.09 tn, where tn is the nominal pipe thickness. This is due to the fact that the current provisions would result in a weld that would be weaker than the VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 pipe to which it is adjoined under these dimensions. The weld stress provisions in the version of the footnotes contained in the 1989 Addenda have been relocated to different subparagraphs in subsequent BPV Code editions and addenda. Therefore, the current Code’s reference in Footnote 11 to Figures NC– 3673.2(b)–1 and ND–3673.2(b)–1 is not correct for BPV Code editions and addenda after the 1989 Addenda, in applying the condition. The proposed rule would correct this issue by clearly identifying the prohibited code provisions in the editions and addenda in a tabular format. As an editorial matter, this proposed rule identifies the prohibited BPV Code provisions as ‘‘notes,’’ which is the term used by the ASME, rather than ‘‘footnotes.’’ The NRC proposes to use the terminology used by the ASME for clarity. 10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance The NRC proposes to approve for use the version of NQA–1 referenced in the 2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code, Section III, Subsection NCA, Article 7000, which this rule is also incorporating by reference. This will allow applicants and licensees to use the 2008 Edition and the 2009–1a Addenda of NQA–1 when using the 2010 and later editions and addenda of Section III. In the 2010 Edition of ASME BPV Code, Section III, Subsection NCA, Article NCA–4000, ‘‘Quality Assurance,’’ was updated to require NType Certificate Holders to comply with the requirements of Part 1 of the 2008 Edition and the 2009–1a Addenda of ASME Standard NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications,’’ as modified and supplemented in NCA–4120(b) and NCA–4134. In addition, NCA–4110(b) was revised to remove the reference to a specific edition and addenda of ASME NQA–1, and Table NCA–7100–2, ‘‘Standards and Specifications Referenced in Division 1,’’ was updated to require the 2008 Edition and 2009– 1a Addenda of NQA–1 when using the 2010 Edition of Section III. The NRC reviewed the 2008 Edition and the 2009–1a Addenda of NQA–1 and compared it to previously approved versions of NQA–1 and found that there were no significant differences. In addition, the NRC reviewed the changes to Subsection NCA that reference the 2008 Edition and 2009–1a Addenda of NQA–1, compared them to previously approved versions of Subsection NCA, and found that there were no significant differences. Therefore, the NRC has PO 00000 Frm 00005 Fmt 4701 Sfmt 4702 56823 concluded that these Editions and Addenda of NQA–1 are acceptable for use. The NRC proposes to revise § 50.55a(b)(1)(iv) to clarify that an applicant’s or licensee’s commitments, addressing those areas where NQA–1 either does not address a requirement in appendix B to 10 CFR part 50, ‘‘Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,’’ or is less stringent than the comparable appendix B requirement, governs the applicant’s or licensee’s Section III activities. The proposed clarification is consistent with § 50.55a(b)(2)(x) and § 50.55a(b)(3)(i). NQA–1 provides the ASME’s method for establishing and implementing a quality assurance (QA) program for the design and construction of nuclear power plants and fuel reprocessing plants. However, NQA–1, as modified and supplemented in NCA– 4120(b) and NCA–4134, does not address some of the requirements of appendix B to 10 CFR part 50. In some cases, the provisions of NQA–1 are less stringent than the comparable appendix B requirement. Thus, in order to meet the requirements of appendix B, an applicant’s or licensee’s QA program description must contain commitments addressing those provisions of appendix B which are not covered by NQA–1, as well as provisions that supplement or replace the NQA–1 provisions where the appendix B requirement is more stringent. Finally, the NRC is considering removing the reference in § 50.55a(b)(1)(iv) to versions of NQA–1 older than the 1994 Edition. The NRC requests public comment on whether any applicant or licensee is committed to, and is using, a version of NQA–1 older than the 1994 Edition, and if so, what version the applicant or licensee is using. 10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification and Demonstration of Function of Incompressible-Fluid Pressure-Relief Valves The NRC proposes to revise § 50.55a(b)(1)(vii) so that the existing condition prohibiting the use of paragraph NB–7742(a)(2) of the 2006 Addenda through the 2007 Edition up to and including the 2008 Addenda is extended to include the editions and addenda up to the 2013 Edition which are the subject of this rulemaking. 10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME Certification Marks The NRC is proposing to add new paragraph, § 50.55a(b)(1)(viii), to allow E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56824 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules licensees to use either the ASME BPV Code Symbol Stamps of editions and addenda earlier than the 2011 Addenda to the 2010 Edition of the ASME BPV Code or the ASME Certification Marks with the appropriate certification designators and class designators as specified in the 2013 Edition through the latest edition and addenda incorporated by reference in 10 CFR 50.55a. The ASME BPV Code requires, in certain instances, that components be stamped. The stamp signifies that the component has been designed, fabricated, examined and tested, as specified in the ASME BPV Code. The stamp also signifies that the required ASME BPV Code data report forms have been completed, and the authorized inspector has inspected the item and authorized the application of the ASME BPV Code Symbol Stamp. The ASME has instituted changes in the BPV Code to consolidate the different ASME BPV Code Symbol Stamps into a common ASME Certification Mark. This action was implemented in the 2011 Addenda to the 2010 Edition of the ASME BPV Code. As of the end of 2012, ASME no longer utilizes the ASME BPV Code Symbol Stamp. Licensees, however, may not have updated to the Edition or Addenda that identifies the use of the ASME Certification Mark. Nevertheless, licensees are legally required to implement the ASME BPV Code Edition and Addenda identified as their current code of record. As ASME components are procured, these components may be received with the ASME Certification Mark, while the licensee’s current code of record may require the component to have the ASME BPV Code Symbol Stamp. Installation of a component under such circumstances would not be in compliance with the regulations that the licensees are required to meet. Both the ASME Certification Mark and the ASME BPV Code Symbol Stamp are official ASME methods of certifying compliance with the Code. Although these ASME Certification Marks differ slightly in appearance, they serve the same purpose of certifying code compliance by the ASME Certificate Holder and continue to provide for the same level of quality assurance for the application of the ASME Certification Mark as was required for the application of the ASME BPV Code Symbol Stamp. The new ASME Certification Mark represents a small, non-safety significant modification of ASME’s trademark. As such, it does not change the technical requirements of the Code. ASME has confirmed that the Certification Mark with designator is VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 equivalent to the corresponding BPV Code Symbol Stamp. Based on statements by ASME in a letter dated August 17, 2012, the NRC has concluded that the ASME BPV Code Symbol Stamps and ASME Certification Mark with code-specific designators are equivalent with respect to their certification of compliance with the BPV Code. The NRC discussed this issue in Regulatory Issue Summary 2013–07, ‘‘NRC Staff Position on the Use of American Society of Mechanical Engineers Certification Mark,’’ dated May 28, 2013. B. ASME BPV Code, Section XI 10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section XI The NRC proposes to revise § 50.55a(a)(1)(ii) to clarify that Section XI Non-mandatory Appendix U of the 2013 Edition of ASME BPV Code Section XI is not incorporated by reference and therefore not approved for use. The NRC is developing an integrated approach to the issue of operational leakage. The NRC has not completed its determination of how Appendix U fits into this integrated approach to address the operational leakage issue at nuclear power plants. The operational leakage issue has many factors that need to be considered such as acceptance criteria, corrective actions, application of repair/ replacement requirements, component operability determination, concerns related to continued operation, maximum acceptable leakage rates, flaw growth rates, flaw measurement techniques, schedules for eliminating leakage, and when or if the leakage requires authorization by the NRC. The NRC plans to complete the development of the regulatory approach to operational leakage and issue it in a future rulemaking. 10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and Addenda of Subsection IWE and Subsection IWL The NRC proposes to revise § 50.55a(b)(2)(vi) to explicitly state that the provision requiring the use of either the 1992 Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and Subsection IWL when implementing the initial 120-month containment inservice inspection program applies only to those licensees that were required by previous versions 2 of § 50.55a to 2 See the supplementary information and rule language for § 50.55a(b)(2)(vi), § 50.55a(g)(4), and § 50.55a(g)(6)(ii)(B) in Federal Register notices PO 00000 Frm 00006 Fmt 4701 Sfmt 4702 develop and implement a containment inservice inspection program in accordance with Subsection IWE and Subsection IWL, and complete an expedited examination of containment during the 5-year period from September 9, 1996, to September 9, 2001. The expedited examination involved the completion of the first set of examinations of the first or initial 120month containment inspection interval. It is noted that all the operating reactors in the above stated class would have gone past their initial 120-month inspection interval by 2011. The proposed change removes the possibility of misinterpretation of the provision as requiring plants that do not fall in the above class, such as reactors licensed after September 9, 2001, to use the 1992 Edition with 1992 Addenda or the 1995 Edition with 1996 Addenda of Subsection IWE and Subsection IWL, Section XI for implementing the initial 120-month inspection interval of the containment inservice inspection program. Applicants and licensees that do not fall in the above class must use Code editions and addenda in accordance with § 50.55a(g)(4)(i) and (g)(4)(ii), respectively, for the initial and successive 120-month containment inservice inspection intervals. 10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment Examinations The NRC proposes to revise § 50.55a(b)(2)(viii) by removing the condition for using the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL requiring compliance with § 50.55a(b)(2)(viii)(E) and adding a requirement to comply with § 50.55a(b)(2)(viii)(H) and (I). Section 50.55a(b)(2)(viii)(E) is one of several conditions that apply to the inservice examination of concrete containments using Subsection IWL of various editions and addenda of the ASME BPV Code, Section XI, incorporated by reference in § 50.55a(a)(1)(ii). The NRC proposes to remove the condition in § 50.55a(b)(2)(viii)(E) when applying the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL because its intent has been incorporated into the Code in the new provision IWL–2512, ‘‘Inaccessible Areas.’’ The reasons for requiring compliance with § 50.55a(b)(2)(viii)(H) and (I) are set forth in the next two sections. published on August 8, 1996 (61 FR 41303), and September 22, 1999 (64 FR 51370). E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules 10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth Provision The NRC proposes to add a new paragraph, § 50.55a(b)(2)(viii)(H), to specify the information that must be provided in the ISI Summary Report required by IWA–6000, when inaccessible concrete surfaces are evaluated under the new code provision IWL–2512. This new condition would replace the existing condition in § 50.55a(b)(2)(viii)(E) when using the 2007 Edition with the 2009 Addenda through the 2013 Edition of Subsection IWL. The existing condition in § 50.55a(b)(2)(viii)(E) of the current rule requires that, for Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas, and provide the evaluation information required by §§ 50.55a(b)(2)(viii)(E)(1), (E)(2), and (E)(3) in the IWA–6000 ISI Summary Report. In the 2009 Addenda Subsection IWL, the ASME revised existing provisions IWL–1220 and IWL–2510 and added new provision IWL–2512 intended to incorporate the condition in § 50.55a(b)(2)(viii)(E) into Subsection IWL. The IWL–2510, ‘‘Surface Examination,’’ was restructured into new paragraphs IWL–2511, ‘‘Accessible Areas,’’ with almost the same provisions as the previous IWL–2510 and IWL– 2512, ‘‘Inaccessible Areas,’’ to be specific to examinations required for accessible areas, and differentiate between those and the new requirements for inaccessible areas. The inaccessible areas addressed by the new IWL–2512 are: (1) Concrete surfaces obstructed by adjacent structures, parts or appurtenances (e.g., generally abovegrade inaccessible areas) and (2) concrete surfaces made inaccessible by foundation material or backfill (e.g., below-grade inaccessible areas). The revised IWL–2511(a) has a new requirement that states that, ‘‘If the Responsible Engineer determines that observed suspect conditions indicate the presence of, or could result in, degradation of inaccessible areas, the requirements of IWL–2512(a) shall be met.’’ The new IWL–2512(a) requires the ‘‘Responsible Engineer’’ to evaluate suspect conditions and specify the type and extent of examinations, if any, required to be performed on inaccessible surface areas described in the previous paragraph. The acceptability of the evaluated VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 inaccessible area would be determined either based on the evaluation or based on the additional examinations, if determined to be required. The new IWL–2512(b) further requires a periodic technical evaluation of below-grade inaccessible areas of concrete to be performed to determine and manage its susceptibility to degradation regardless of whether suspect conditions exist in accessible areas that would warrant an evaluation of inaccessible areas based on the condition in § 50.55a(b)(2)(viii)(E). Therefore, the revised IWL–2511(a) and new IWL– 2512 code provisions address the evaluation and acceptability of inaccessible areas consistent with the existing condition in § 50.55a(b)(2)(viii)(E), with one exception. The exception is that the new IWL–2512 provision does not explicitly require the information specified in §§ 50.55a(b)(2)(viii)(E)(1), (E)(2), and (E)(3) of the existing condition to be provided in the IWA–6000 ISI Summary Report. For these reasons, the NRC proposes to identify the information that must be provided in the ISI Summary Report required by IWA–6000 when inaccessible concrete surfaces are evaluated under the new code provision IWL–2512. This new condition would replace the existing condition in § 50.55a(b)(2)(viii)(E) when using the 2007 Edition with the 2009 Addenda through the 2013 Edition of Subsection IWL. The information requested by the new condition must be provided when inaccessible concrete areas are evaluated per IWL–2512(a) for degradation based on suspect conditions found in accessible areas, as well as when periodic technical evaluations of inaccessible below-grade concrete areas required by IWL–2512(b) are performed. 10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth Provision The NRC proposes to add § 50.55a(b)(2)(viii)(I) to place a condition on the periodic technical evaluation requirements in the new IWL–2512(b), for consistency with NUREG–1801, Revision 2, ‘‘Generic Aging Lessons Learned (GALL) Report,’’ with regard to aging management of below-grade containment concrete surfaces. The new IWL–2512(b) provision is applicable to inaccessible below-grade concrete surfaces exposed to foundation soil, backfill, or groundwater. This condition would apply only during the period of extended operation of a renewed license under 10 CFR part 54, when using IWL– 2512(b) of the 2007 Edition with 2009 PO 00000 Frm 00007 Fmt 4701 Sfmt 4702 56825 Addenda through the 2013 Edition of Subsection IWL. In the 2009 Addenda of Subsection IWL, the ASME added new code provisions, IWL–2512(b) and (c) as well as a new line item L1.13 in Table IWL– 2500–1, intended to specifically address aging management concerns with potentially unidentified degradation of inaccessible below-grade containment concrete areas and to be responsive to actions outlined in the GALL Report related to aging management of inaccessible below-grade concrete surfaces. It is noted that these new code provisions are an enhancement to the requirement of the existing condition in § 50.55a(b)(2)(viii)(E) to specifically address aging management of inaccessible below-grade containment concrete areas and is generally acceptable to the NRC. The new IWL–2512(b) provides requirements for systematically performing a periodic technical evaluation of concrete surfaces exposed to foundation soil, backfill, or groundwater to determine susceptibility of the concrete to deterioration that could affect its ability to perform its intended design function under conditions anticipated through the service life of the structure. It requires the technical evaluation to be performed and documented at periodic intervals not to exceed 10 years regardless of whether conditions exist in accessible areas that would warrant an evaluation of inaccessible areas by the existing condition in § 50.55a(b)(2)(viii)(E), which the NRC finds reasonable for the initial 40-year operating license period. The new IWL–2512(b) further provides the specific elements, including aging mechanisms considered, that the technical evaluation should include, as well as the definition of an aggressive below-grade environment. The new IWL–2512(c) requires that the evaluation results of IWL–2512(b) be used to define and document the condition monitoring program, if determined to be required, including required examinations and frequencies, to be implemented for the management of degradation and aging effects of the below-grade concrete surface areas. If it is determined that additional examinations are required, these examinations of inaccessible belowgrade areas will be implemented in accordance with new line item L1.13 in Table IWL–2500–1 under Examination Category L–A, Concrete, with acceptance criteria based on IWL–3210. It should be noted that a technical evaluation approach, such as in IWL– 2512(b), could be used, and is generally used, to determine acceptability of a E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56826 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules below-grade inaccessible area to satisfy the condition in § 50.55a(b)(2)(viii)(E). The technical evaluation requirements in IWL–2512(b) help to determine the susceptibility to degradation and manage aging effects of inaccessible below-grade concrete surfaces, before the loss of intended function. The requirements are based on, and are generally consistent with, the guidance in the GALL Report,’’ with the following two exceptions. The first exception is that IWL–2512(b) requires the technical evaluation to determine the susceptibility of the concrete to degradation and the ability to perform the intended design function through its service life at periodic intervals not to exceed 10 years. The aging management programs (AMPs) for safety-related structures (e.g., Structures Monitoring) in the GALL Report require such evaluation to be performed at intervals not to exceed 5 years, which is also consistent with applicant commitments during review of license renewal applications. The second exception is that IWL–2512(b) requires that examination of representative samples of below-grade concrete be performed if excavated for any reason when an aggressive below-grade environment is present. However, the AMPs (X1.S6 Structures Monitoring and X1.S7 Water Control Structures) in the GALL Report require the same examination even for a non-aggressive below-grade environment. Based on these reasons, the NRC proposes to add a new § 50.55a(b)(2)(viii)(I) to place a condition on the periodic technical evaluation requirements in IWL–2512(b) for consistency with the GALL Report, with regard to aging management of inaccessible below-grade concrete components of the containment. The new IWL–2512(b) is applicable to inaccessible below-grade concrete surfaces of the containment cylindrical wall and basemat foundations, which are exposed to foundation soil, backfill, or groundwater. The new condition requires that, during the period of extended operation of a renewed license, the technical evaluation under IWL–2512(b) of inaccessible belowgrade concrete surfaces exposed to foundation soil, backfill, or groundwater be performed at periodic intervals not to exceed 5 years. Also, the condition requires the examination of representative samples of the exposed portions of the below-grade concrete be performed when excavated for any reason. Since the GALL Report is the technical basis document for license renewal, this new condition applies only during the period of extended VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 operation of a renewed license under 10 CFR part 54, when using IWL–2512(b) of the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL, Section XI. 10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment Examinations The NRC proposes to continue to apply the existing conditions in §§ 50.55a(b)(2)(ix)(A)(2), (b)(2)(ix)(B), and (b)(2)(ix)(J) governing examinations of metal containments and the liners of concrete containments under Subsection IWE to the 2007 Edition with 2009 Addenda through the 2013 Edition (the code editions and addenda which are the subject of this rulemaking). The NRC reviewed the code changes in Subsection IWE of the 2009 Addenda through the 2013 Edition of ASME BPV Code, Section XI, and notes that all of the changes were editorial or administrative with the intent to improve the clarity of the existing requirements or correct errors by errata. There were no changes to Subsection IWE in the code editions and addenda that are the subject of this rulemaking that the NRC believes would require new regulatory conditions to ensure safety, nor do the changes to Subsection IWE address the NRC’s reasons for adopting the conditions on the use of Subsection IWE. Although this continuation of the applicability of the three conditions does not require a rule change, the NRC is discussing this for the benefit of stakeholder understanding of the effect of the proposed rule. 10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance The NRC proposes to approve for use the version of NQA–1 referenced in the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition of the ASME BPV Code, Section XI, Table IWA 1600–1, ‘‘Referenced Standards and Specifications,’’ which this rule is also incorporating by reference. This will allow licensees to use the 1994 or the 2008 Edition and the 2009–1a Addenda of NQA–1 when using the 2009 Addenda and later editions and addenda of Section XI. In the 2013 Edition of ASME BPV Code, Section XI, Table IWA 1600–1 was updated to allow licensees to use the 1994 or the 2008 Edition with the 2009–1a Addenda of NQA–1 when using the 2013 Edition of Section XI. In the 2010 Edition of ASME BPV Code, Section XI, IWA–1400, ‘‘Owner’s Responsibilities,’’ subparagraph (n)(2) was updated to reference the NQA–1 Part I, Basic Requirements and PO 00000 Frm 00008 Fmt 4701 Sfmt 4702 Supplementary Requirements for Nuclear Facilities. In the 2009 Addenda of the 2007 Edition of ASME BPV Code, Section XI, Table IWA–1600–1, ‘‘Referenced Standards and Specifications,’’ was updated to allow licensees to use the 1994 Edition of NQA–1. The NRC reviewed the 2008 Edition and the 2009–1a Addenda of NQA–1 and compared it to previously approved versions of NQA–1 and found that there were no significant differences. Therefore, the NRC has concluded that these Editions and Addenda of NQA–1 are acceptable for use. The NRC proposes to amend § 50.55a(b)(2)(x) to clarify that a licensee’s commitments addressing those areas where NQA–1 either does not address an appendix B requirement or is less stringent than the comparable appendix B requirement governs the licensee’s Section XI activities. The proposed clarification is consistent with §§ 50.55a(b)(1)(iv) and (b)(3)(i). The ASME’s method for establishing and implementing a QA program for the design and construction of nuclear power plants and fuel reprocessing plants is described in NQA–1. However, NQA–1 does not address some of the requirements of appendix B to 10 CFR part 50. In some cases, the provisions of NQA–1 are less stringent than the comparable appendix B requirement. Thus, in order to meet the requirements of appendix B, a licensee’s QA program description must contain commitments addressing those provisions of appendix B which are not covered by NQA–1, as well as provisions that supplement or replace the NQA–1 provisions where the appendix B requirement is more stringent. Finally, the NRC is considering removing the reference in § 50.55a(b)(2)(x) to versions of NQA–1 older than the 1994 Edition. The NRC requests public comment on whether any licensee is committed to, and is using, a version of NQA–1 older than the 1994 Edition, and if so, what version the applicant or licensee is using. 10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth Provision The NRC proposes to add a new paragraph, § 50.55a(b)(2)(xviii)(D), to prohibit applicants and licensees from using the ultrasonic examination nondestructive examination (NDE) personnel certification requirements in Section XI, Appendix VII and subarticle VIII–2200 of the 2011 Addenda and 2013 Edition of the ASME BPV Code. Section 50.55a(b)(2)(xviii) currently includes conditions on the certification E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 of NDE personnel. In addition, the new paragraph would require applicants and licensees to use the 2010 Edition, Table VII–4110–1 training hour requirements for Levels I, II, and III ultrasonic examination personnel, and the 2010 Edition, subarticle VIII–2200 of Appendix VIII prerequisites for personnel requirements. In the 2011 Addenda and 2013 Edition, the ASME BPV Code added an accelerated Appendix VII training process for certification of ultrasonic examination personnel based on training and prior experience, and separated the Appendix VII training requirements from the Appendix VIII qualification requirements. These new ASME BPV Code provisions would provide personnel in training with less experience and exposure to representative flaws in representative materials and configurations common to operating nuclear power plants, and they would permit personnel with prior non-nuclear ultrasonic examination experience to qualify for examinations in nuclear power plants without exposure to the variety of defects, examination conditions, components, and regulations common to operating nuclear power plants. The impact of reduced training and nuclear power plant familiarization is unknown. The ASME BPV Code supplants training hours and field experience without a technical basis, minimum defined training criteria, process details, or standardization. For these reasons, the NRC is proposing to prohibit the use of Appendix VII and VIII–2200 in the 2011 Addenda and 2013 Edition, and instead require applicants and licensees using the 2011 Addenda and 2013 Edition to use Table VII–4110–1 in the 2010 Edition, and VIII–2200, Appendix VIII prerequisites for ultrasonic examination personnel requirements in the 2010 Edition. 10 CFR 50.55a(b)(2)(xxi)(A) Table IWB– 2500–1 Examination Requirements: First Provision The NRC proposes to revise § 50.55a(b)(2)(xxi)(A) to modify the standard for visual magnification resolution sensitivity and contrast for visual examinations performed on Examination Category B–D components instead of ultrasonic examinations, making the rule conform with ASME BPV Code, Section XI requirements for VT–1 examinations. The character recognition rules are used in ASME BPV Code, Section XI, Table IWA–2211–1 for VT–1 tests, and are the standard tests used for resolution and contrast checks of VT–1 equipment. This revision essentially removes a requirement that VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 was in addition to ASME BPV Code that required 1-mil wires to be used in licensees’ Sensitivity, Resolution and Contrast Standard targets. In 2004, the NRC published NUREG/CR–6860, ‘‘An Assessment of Visual Testing,’’ showing that a linear target, such as a wire, is not an effective method for testing the resolution of a video camera system. In addition, BWRVIP–03 was changed to eliminate a 1⁄2 mil wire from the Sensitivity Resolution and Contrast Standards due to similar concerns. Simple line detection can be a poor performance standard, allowing detection of a highly blurred image. This does not emulate sharpness quality recognition for evaluation of weld discontinuities. The 750 mm (30 mil) and the even smaller 25 mm (1 mil) widths should not be used as performance standards because they do not determine image sharpness. This technique only measures the ‘‘visible minimum’’ for long linear indications, and does not measure a system’s resolution or recognition limits. If the wire, or printed line, has a strong enough contrast against the background, then a linear feature well below the resolution of a system can be detected. 10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator Preservice Examinations The NRC proposes to add § 50.55a(b)(2)(xxx) to require a full length examination of 100 percent of the tubing in each newly installed steam generator prior to plant startup. This requirement would be instead of the unapproved provisions in IWB–2200(c) pertaining to steam generator tube preservice inspections (PSI). Steam generator tubes, a significant portion of the reactor coolant pressure boundary, are important to the safe operation of a pressurized water reactor. As such, the NRC has established requirements pertaining to the design, fabrication, erection, testing, and inspection of the steam generator tubes. With respect to the performance of the PSI of steam generator tubes, the NRC has indicated in NRC Regulatory Guide (RG) 1.83, Revision 1, ‘‘Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes,’’ (withdrawn in 2009) that all tubes in the steam generator should be inspected by eddy current or alternative technique prior to service to establish a baseline condition of the tubing. A similar position is articulated in NUREG–0800, Standard Review Plan (SRP) Section 5.4.2.2, ‘‘Steam Generator Tube Inservice Inspection,’’ Revision 1 and subsequent revisions. A PSI is important since it ensures that the steam generator tubes PO 00000 Frm 00009 Fmt 4701 Sfmt 4702 56827 are acceptable for initial operation. In addition, the PSI provides the baseline condition of the tubes. This data is essential in assessing the nature of indications found in the tubes during subsequent inservice inspections. Preservice requirements for ASME Class 1 components are provided in IWB–2200, and IWB–2200(c) currently states, ‘‘Steam generator tube examination shall be governed by the plant Technical Specifications (TS).’’ However, there are no preservice examination requirements for steam generators defined in plant TS. Preservice examination requirements for steam generators are not within any of the categories described in 50.36 for the content of TS. Because IWB–2200(c) requires the steam generator tube examinations be performed in accordance with plant TS, and TS contain no rules for PSI of steam generator tubing, the NRC is clarifying the preservice inspection requirements for steam generator tubes. The proposed clarification is consistent with industry guidelines and the NRC staff position outlined in SRP Section 5.4.2.2, ‘‘Steam Generator Program.’’ The proposed requirement supersedes the requirements of IWB– 2200(c). These inspections must be performed with the objective of finding and characterizing the types of preservice flaws that may be present in the tubes and flaws that may occur during operation. 10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping Devices The NRC proposes to add § 50.55a(b)(2)(xxxi) to prohibit the use of mechanical clamping devices on Class 1 piping and portions of piping systems that form the containment boundary. In the 2010 Edition of the ASME BPV Code, a change was made to include mechanical clamping devices under the small items exclusion rules of IWA–4131. Currently in the 2007 Edition/2008 Addenda of Section XI under IWA–4133, ‘‘Mechanical Clamping Devices Used as Piping Pressure Boundary,’’ mechanical clamping devices may be used only if they meet the requirements of Mandatory Appendix IX of Section XI of the ASME BPV Code. Article IX–1000 (c) of Appendix IX prohibits the use of mechanical clamping devices on (1) Class 1 piping and (2) portions of a piping system that form the containment boundary. In the 2010 Edition, IWA–4133 was modified to allow use of IWA–4131.1(c) for the installation of mechanical clamping devices. This change allowed E:\FR\FM\18SEP5.SGM 18SEP5 56828 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 the use of small items exemption rules in the installation of mechanical clamps. Subparagraph IWA–4131.1(c) was added such that mechanical clamping devices installed on items classified as ‘‘small items’’ under IWA– 4131, including Class 1 piping and portions of a piping system that form the containment boundary, would be allowed without a repair/replacement plan, pressure testing, services of an Authorized Inspection Agency, and completion of NIS–2 form. The NRC, in accordance with the previously approved IWA–4133 of the 2007 Edition/2008 Addenda of the ASME BPV Code, does not believe that the ASME has provided a sufficient technical basis to support the use of mechanical clamps on Class 1 piping or portions of a piping system that form the containment boundary as a permanent repair. Furthermore, the NRC does not believe that the ASME has provided any basis for the small item exemption allowing the installation of mechanical clamps on these components. In the 2011 Addenda of the ASME BPV Code, IWA–4131.1(c) was relocated to IWA–4131.1(d). 10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report Submittal The NRC proposes to add § 50.55a(b)(2)(xxxii) to require licensees using the 2010 Edition and later editions and addenda of Section XI to continue to submit Summary Reports as required in IWA–6240 of the 2009 Addenda. Prior to the 2010 Edition, Section XI required the preservice summary report to be submitted prior to the date of placement of the unit into commercial service, and the inservice summary report to be submitted within 90 calendar days of the completion of each refueling outage. In the 2010 Edition, IWA–6240 was revised to state, ‘‘Summary Reports shall be submitted to the enforcement and regulatory authorities having jurisdiction at the plant site, if required by these authorities.’’ This change in the 2010 Edition could lead to confusion as to whether or not the summary reports need to be submitted to the NRC, as well as the time for submitting the reports if they were required. The NRC believes that summary reports must continue to be submitted to the NRC in a timely manner because they provide valuable information regarding examinations performed, conditions noted, corrective actions taken, and the implementation status of PSI and ISI programs. Therefore, the NRC proposes adding § 50.55a(b)(2)(xxxii) to ensure that preservice and inservice summary VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 reports will continue to be submitted within the timeframes currently established in Section XI editions and addenda prior to the 2010 Edition. 10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed Allowable Pressure The NRC proposes to add § 50.55a(b)(2)(xxxiii) to prohibit the use of Appendix G Paragraph G–2216 in the 2011 Addenda and later editions and addenda of the ASME BPV Code, Section XI. The 2011 Addenda of the ASME BPV Code included, for the first time, a risk-informed methodology to compute allowable pressure as a function of inlet temperature for reactor heat-up and cool-down at rates not to exceed 100 degrees F/hr (56 degrees C/ hr). This methodology was developed based upon probabilistic fracture mechanics (PFM) evaluations that investigated the likelihood of reactor pressure vessel (RPV) failure based on specific heat-up and cool-down scenarios. During the ASME’s consideration of this change, the NRC staff noted that additional requirements would need to be placed on the use of this alternative. For example, the NRC staff indicated that it would be important for a licensee who wishes to utilize such a riskinformed methodology for determining plant-specific pressure-temperature limits to ensure that the material condition of its facility is consistent with assumptions made in the PFM evaluations that supported the development of the methodology. One aspect of this would be evaluating plantspecific inservice inspection data to determine whether the facility’s RPV flaw distribution was consistent with the flaw distribution assumed in the supporting PFM evaluations. This consideration is consistent with a similar requirement established by the NRC in § 50.61a, ‘‘Alternative Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events.’’ The PFM methodology that supports § 50.61a is very similar that which was used to support ASME BPV Code, Section XI, Appendix G, Paragraph G–2216. These concerns with the Paragraph G–2216 methodology for computing allowable pressure as a function of inlet temperature for reactor heat-up and cooldown were not addressed by the ASME. Accordingly, the NRC is proposing to prohibit the use of Paragraph G–2216 in Appendix G of the 2010 Edition. The continued use of the deterministic methodology of Section XI, Appendix G to generate P– T limits remains acceptable. PO 00000 Frm 00010 Fmt 4701 Sfmt 4702 10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Disposition of Flaws in Class 3 Components The NRC proposes to add § 50.55a(b)(2)(xxxiv) to require that when using the 2013 Edition of the ASME BPV Code, Section XI, the licensee shall use the acceptance standards of IWD–3510 for the disposition of flaws in Category D–A components (i.e., welded attachments for vessels, piping, pumps, and valves). The 2013 Edition of the ASME BPV Code, Section XI, IWD–3510, ‘‘Standards for Examination Category D– A, Welded Attachments for Vessels, Piping, Pumps, and Valves,’’ states that the acceptance standards are: ‘‘In the course of preparation, the requirements of IWC–3500 may be used.’’ The ASME BPV Code, Section XI, IWD–3410, ‘‘Acceptance Standards,’’ states that the acceptance standards referenced in Table IWD–3410–1 shall be applied to determine acceptability for service. Table IWD–3410–1 states that the acceptance standard for Examination Category D–A is IWB–3510. A discrepancy exists between the provisions in IWD–3410, which references Table IWD–3410–1, and the provisions in IWD–3510. The provisions in IWD–3510 require the use of the acceptance standards of IWC–3500 whereas Table IWD–3410–1 requires the use of the acceptance standards of IWB– 3510 to disposition flaws detected in the Examination Category D–A components. Both IWD–3410 and IWD–3510 should reference the same subarticle and acceptance standards. The NRC believes that this discrepancy is due to an error in the publishing of the 2013 Edition because the code committee action which proposed the revised Class 3 acceptance criteria and added Table IWD–3410–1 showed the appropriate Acceptance Standard to be IWD–3510. The intent of the condition is to provide clarification and consistency in requirements between IWD–3410 and IWD–3510. 10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 in the KIa and KIc Equations The NRC proposes to add § 50.55a(b)(2)(xxxv) to specify that when licensees use the 2013 Edition of the ASME BPV Code, Section XI, Appendix A, paragraph A–4200, if T0 is available, then RTT0 may be used in place of RTNDT for applications using the KIc equation and the associated KIc curve, but not for applications using the KIa equation and the associated KIa curve. Non-mandatory Appendix A provides a procedure based on linear elastic E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules fracture mechanics (LEFM) for determining the acceptability of flaws that have been detected during inservice inspections that exceed the allowable flaw indication standards of IWB–3500. Sub-article A–4200 provides a procedure for determining fracture toughness of the material used in the LEFM analysis. The NRC staff’s concern is related to the proposed insertion regarding an alternative based on Master Curve methodology to determine the nil-ductility transition reference temperature RTNDT, which is an important parameter in determining the fracture toughness of the material. Specifically, the insertion proposed to use Master Curve reference temperature RTT0, which is defined as RTT0 = T0 + 35 °F, where T0 is a material-specific temperature value determined in accordance with ASTM E1921, ‘‘Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range,’’ to index (shift) the fracture toughness KIc curve, based on the lower bound of static initiation critical stress intensity factor, as well as the KIa curve, based on the lower bound of crack arrest critical stress intensity factor. While use of RTT0 to index the KIc curve is acceptable, using RTT0 to index the KIa curve is questionable. This NRC staff concern is based on the data analysis in ‘‘A Physics-Based Model for the Crack Arrest Toughness of Ferritic Steels,’’ written by NRC staff member Mark Kirk, and published in ‘‘Fatigue and Fracture Mechanics, 33rd Volume, ASTM STP 1417,’’ which indicated that the crack arrest data does not support using RTT0 as RTNDT to index the KIa curve. This is also confirmed by industry data disclosed in a presentation, ‘‘Final Results from the CARINA Project on Crack Initiation and Arrest of Irradiated German RPV Steels for Neutron Fluences in the Upper Bound,’’ by AREVA at the 26th Symposium on Effects of Radiation on Nuclear Materials (June 12–13, 2013, Indianapolis, IN, USA). The NRC staff recognized that the proposed insertion is consistent with Code Case N–629, ‘‘Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials,’’ which was accepted by the NRC without conditions. In addition to the current NRC effort, the appropriate ASME Code committee is in the process of correcting this issue in a future revision of Appendix A of Section XI. With this condition, users of Appendix A can avoid using an erroneous fracture toughness KIa value in their LEFM analysis for determining VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 the acceptability of a detected flaw in applicable components. Therefore, the NRC is proposing to add a condition which permits the use of RTT0 in place of RTNDT in applications using the KIc equation and the associated KIc curve, but does not permit the use of RTT0 in place of RTNDT in applications using the KIa equation and the associated KIa curve. 10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of Irradiated Materials The NRC proposes to add § 50.55a(b)(2)(xxxvi) to require licensees using ASME BPV Code, Section XI, 2013 Edition, Appendix A, paragraph A–4400, to obtain NRC approval before using irradiated T0 and the associated RTT0 in establishing fracture toughness of irradiated materials. Sub-article A–4400 provides guidance for considering irradiation effects on materials. The NRC staff’s concern is related to use of RTT0 based on measured T0 of the irradiated materials. Specifically, the NRC staff has concerns over this sentence in the proposed insertion: ‘‘Measurement of RTT0 of unirradiated or irradiated materials as defined in A–4200(b) is permitted, including use of the procedures given in ASTM E1921 to obtain direct measurement of irradiated T0.’’ Permission of measurement of RTT0 of irradiated materials, without providing guidelines regarding how to use the measured parameter in determining the fracture toughness of the irradiated materials, may mislead the users of Appendix A into adopting methodology not accepted by the NRC. With this condition, users of Appendix A can avoid using a fracture toughness KIc value based on the irradiated T0 and the associated RTT0 in their LEFM analysis for determining the acceptability of a detected flaw in applicable components. 10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements The NRC proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and (g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i), (g)(3)(ii), and (g)(3)(v) to distinguish the requirements for accessibility and preservice examination from those for inservice inspection in § 50.55a(g). No substantive change to the requirements is intended by these revisions. C. ASME OM Code 10 CFR 50.55a(b)(3) Conditions on ASME OM Code The NRC proposes to revise § 50.55a(b)(3) to clarify that Subsections PO 00000 Frm 00011 Fmt 4701 Sfmt 4702 56829 ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and Non-mandatory Appendices A through H and J through M of the ASME OM Code would be incorporated by reference in § 50.55a. The NRC is clarifying that the ASME OM Code nonmandatory appendices, which are incorporated by reference into § 50.55a are approved for use, but are not mandated. The non-mandatory appendices may be used by applicants and licensees of nuclear power plants, subject to the conditions in § 50.55a(b)(3). 10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance The NRC proposes to revise § 50.55a(b)(3)(i) to allow use of the 1983 Edition through the 1994 Edition, 2008 Edition, and the 2009–1a Addenda of NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications.’’ The NRC reviewed these Editions and Addenda after the 1983 Edition and compared them to the previously approved versions of NQA– 1 and found that there were no significant differences. The NRC is considering removing the reference in § 50.55a(b)(3)(i) to versions of NQA–1 older than the 1994 Edition. The NRC requests public comment on whether any licensee is committed to, and is using, a version of NQA–1 older than the 1994 Edition and, if so, what version the applicant or licensee is using. 10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) Testing The NRC proposes to revise § 50.55a(b)(3)(ii) to reflect the new Appendix III, ‘‘Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in LightWater Reactor Power Plants,’’ of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. Appendix III of the ASME OM Code establishes provisions for periodic verification of the design-basis capability of MOVs within the scope of the IST program. Appendix III of the ASME OM Code reflects the incorporation of ASME OM Code Cases OMN–1, ‘‘Alternative Rules for Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants,’’ and OMN–11, ‘‘RiskInformed Testing for Motor-Operated Valves.’’ The NRC proposes to add four conditions in new §§ 50.55a(b)(3)(ii)(A), (B), (C), and (D) to address periodic verification of MOV design-basis capability. These conditions are discussed in the next four sections. E:\FR\FM\18SEP5.SGM 18SEP5 56830 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules 10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval The NRC proposes to add § 50.55a(b)(3)(ii)(A) to require that licensees evaluate the adequacy of the diagnostic test interval for each MOV and adjust the interval as necessary, but not later than 5 years or three refueling outages (whichever is longer) from initial implementation of ASME OM Code, Appendix III. Paragraph III– 3310(b) in Appendix III includes a provision stating that if insufficient data exist to determine the IST interval, then MOV inservice testing shall be conducted every two refueling outages or 3 years (whichever is longer) until sufficient data exist, from an applicable MOV or MOV group, to justify a longer IST interval. As discussed in 64 FR 51386 (September 22, 1999) with respect to the use of ASME OM Code Case OMN–1, the NRC considers it appropriate to include a modification requiring licensees to evaluate the information obtained for each MOV, during the first 5 years or three refueling outages (whichever is longer) of the use of Appendix III to validate assumptions made in justifying a longer test interval. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact On Risk The NRC proposes to add § 50.55a(b)(3)(ii)(B) to require that licensees ensure that the potential increase in core damage frequency (CDF) and large early release frequency (LERF) associated with the extension is acceptably small when extending exercise test intervals for high risk MOVs beyond a quarterly frequency. As discussed in 64 FR 51386 (September 22, 1999) with respect to the use of ASME OM Code Case OMN–1, the NRC considers it important for licensees to have sufficient information from the specific MOV, or similar MOVs, to demonstrate that exercising on a refueling outage frequency does not significantly affect component performance. The information may be obtained by grouping similar MOVs and establishing periodic exercising intervals of MOVs in the group over the refueling interval. Section 50.55a(b)(3)(ii)(B) requires that the increase in the overall plant CDF and LERF resulting from the extension be acceptably small. As presented in RG 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on PlantSpecific Changes to the Licensing Basis,’’ the NRC considers acceptably small changes to be relative and to depend on the current plant CDF and LERF. For plants with total baseline VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 CDF of 10¥4 per year or less, acceptably small means CDF increases of up to 10¥5 per year and for plants with total baseline CDF greater than 10¥4 per year, acceptably small means CDF increases of up to 10¥6 per year. For plants with total baseline LERF of 10¥5 per year or less, acceptably small LERF increases are considered to be up to 10¥6 per year, and for plants with total baseline LERF greater than 10¥5 per year, acceptably small LERF increases are considered to be up to 10¥7 per year. 10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization The NRC proposes to add § 50.55a(b)(3)(ii)(C) to require, when applying Appendix III to the ASME OM Code, that licensees categorize MOVs according to their safety significance using the methodology described in ASME OM Code Case OMN–3, ‘‘Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power Plants,’’ subject to the conditions discussed in RG 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety evaluation. Paragraph III–3720 in Appendix III to the ASME OM Code states that when applying risk insights, each MOV shall be evaluated and categorized using a documented risk ranking methodology. Further, Appendix III only addresses risk ranking methodologies that include two risk categories. In light of the potential extension of quarterly test intervals for high risk MOVs and the relaxation of IST activities for low risk MOVs based on risk insights, the NRC has determined that the rule should specify that risk ranking methodologies must have been accepted by the NRC through RG 1.192 (which accepts ASME OM Code Case OMN–3 with the specified conditions) or safety evaluations issued to address plant-specific or industrywide risk ranking methodologies. Two conditions that were previously in RG 1.192 on the use of ASME OM Code Case OMN–11 related to application of the test interval criteria and grouping for low safety significant MOVs have been incorporated in an acceptable manner in Appendix III to the ASME OM Code. As noted in RG 1.192 on the use of ASME OM Code Case OMN–1, the benefits of performing a particular test should be balanced against the potential adverse effects placed on the valves or systems caused by this testing. PO 00000 Frm 00012 Fmt 4701 Sfmt 4702 10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time The NRC proposes to add § 50.55a(b)(3)(ii)(D) to require that when a licensee applies Paragraph III–3600, ‘‘MOV Exercising Requirements,’’ of Appendix III to the OM Code, the licensee verify that the stroke time of the MOV satisfies the assumptions in the plant safety analyses. Previous editions and addenda of the ASME OM Code specified that the licensee must perform quarterly MOV stroke time measurements that could be used to verify that the MOV stroke time satisfies the assumptions in the safety analyses consistent with plant TS. The need for verification of the MOV stroke time during periodic exercising is consistent with the NRC’s lessons learned from the implementation of ASME OM Code Case OMN–1. However, Paragraph III– 3600 of Appendix III of the versions of the OM Code proposed to be incorporated by reference in this rulemaking no longer require the verification of MOV stroke time during periodic exercising. For this reason, the NRC is proposing to adopt the new condition which will effectively retain the need to verify MOV stroke time during periodic exercising. 10 CFR 50.55a(b)(3)(iii) OM condition: New Reactors The NRC proposes to add § 50.55a(b)(3)(iii) to apply specific conditions for IST programs applicable to licensees of new nuclear power plants in addition to the provisions of the ASME OM Code as incorporated by reference with conditions in § 50.55a. Licensees of ‘‘new reactors’’ are, as identified in the proposed paragraph: (i) Holders of operating licenses for nuclear power reactors that received construction permits under this part on or after the date 12 months after the effective date of this rulemaking and (ii) holders of combined licenses (COLs) issued under 10 CFR part 52, whose initial fuel loading occurs on or after the date 12 months after the effective date of this rulemaking. This implementation schedule for new reactors is consistent with the NRC regulations in § 50.55a(f)(4)(i). The NRC is evaluating COL applications to construct and operate nuclear power plants with certified designs under the process described in 10 CFR part 52. Commission Papers SECY–90–016, ‘‘Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements;’’ SECY–93– 087, ‘‘Policy, Technical, and Licensing Issues Pertaining to Evolutionary and E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Advanced Light-Water Reactor (ALWR) Designs;’’ SECY–94–084, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs;’’ and SECY–95–132, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY–94–084),’’ discuss IST programs for new reactors licensed under 10 CFR part 52. In recognition of new reactor designs and lessons learned from nuclear power plant operating experience, the ASME is updating the OM Code to incorporate improved IST provisions for components used in nuclear power plants that were issued (or will be issued) construction permits, or COLs, on or following January 1, 2000 (defined in the ASME OM Code as post-2000 plants). The first phase of the ASME effort incorporated IST provisions that specify full flow pump testing and other clarifications for post-2000 plants in the ASME OM Code beginning with the 2011 Addenda. The second phase of the ASME effort incorporated preservice and inservice inspection and surveillance provisions for pyrotechnicactuated (squib) valves in the 2012 Edition of the ASME OM Code. The ASME is considering further modifications to the ASME OM Code to address additional lessons learned from valve operating experience and new reactor issues. As described in the following paragraphs, § 50.55a(b)(3)(iii) will include four specific conditions. 10 CFR 50.55a(b)(3)(iii)(A) PowerOperated Valves The NRC proposes to add § 50.55a(b)(3)(iii)(A) to require that licensees subject to § 50.55a(b)(3)(iii) develop a program to periodically verify the capability of power-operated valves (POVs) to perform their design-basis safety functions. While Appendix III to the ASME OM Code addresses this requirement for motor-operated valves (MOVs) with applicable conditions specified in § 50.55a, nuclear power plant licensees will need to develop programs to periodically verify the design-basis capability of other POVs. The NRC’s Regulatory Issue Summary (RIS) 2000–03, ‘‘Resolution of Generic Issue 158: Performance of Safety-Related Power-Operated Valves Under Design Basis Conditions,’’ provides attributes for a successful long-term periodic verification program for POVs by incorporating lessons learned from MOV performance at operating nuclear power plants and during research programs. Implementation of Appendix III to the ASME OM Code as accepted VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 in § 50.55a(b)(3)(ii) is acceptable in satisfying § 50.55a(b)(3)(iii)(A) for MOVs. 10 CFR 50.55a(b)(3)(iii)(B) Check Valves The NRC proposes to add § 50.55a(b)(3)(iii)(B) to require that licensees subject to § 50.55a(b)(3)(iii) perform bi-directional testing of check valves within the IST program where practicable. Nuclear power plant operating experience has revealed that testing check valves in only the flow direction can result in significant degradation, such as a missing valve disc, not being identified by the test. Nonmandatory Appendix M, ‘‘Design Guidance for Nuclear Power Plant Systems and Component Testing,’’ to ASME OM Code, 2011 Addenda and 2012 Edition, includes guidance for the design of new reactors to enable bidirectional testing of check valves. New reactor designs will provide the capability for licensees of new nuclear power plants to perform bi-directional testing of check valves within the IST program. 10 CFR 50.55a(b)(3)(iii)(C) FlowInduced Vibration The NRC proposes to add § 50.55a(b)(3)(iii)(C) to require that licensees subject to § 50.55a(b)(3)(iii) monitor flow-induced vibration (FIV) from hydrodynamic loads and acoustic resonance during preservice testing and inservice testing to identify potential adverse flow effects that might impact components within the scope of the IST program. Nuclear power plant operating experience has revealed the potential for adverse flow effects from vibration caused by hydrodynamic loads and acoustic resonance on components in the reactor coolant, steam, and feedwater systems. Therefore, the licensee will need to address potential adverse flow effects on safety-related pumps, valves, and dynamic restraints within the IST program in the reactor coolant, steam, and feedwater systems from hydraulic loading and acoustic resonance during plant operation to confirm that piping, components, restraints, and supports have been designed to withstand the dynamic effects of steady-state FIV and anticipated operational transient conditions. The initial test program can be used to verify that safety-related piping and components are properly installed and supported such that vibrations caused by steady-state or dynamic effects do not result in excessive stress or fatigue in safetyrelated plant systems. PO 00000 Frm 00013 Fmt 4701 Sfmt 4702 56831 10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems The NRC proposes to add § 50.55a(b)(3)(iii)(D) to require that licensees subject to § 50.55a(b)(3)(iii) establish a program to assess the operational readiness of pumps, valves, and dynamic restraints within the scope of the Regulatory Treatment of NonSafety Systems (RTNSS) for applicable reactor designs. In SECY–94–084 and SECY–95–132, the Commission discusses RTNSS policy and technical issues associated with passive plant designs. Some new nuclear power plants have ALWR designs that use passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy, to supply safety injection water and to provide reactor core and containment cooling. Active systems in passive ALWR designs are categorized as nonsafety systems with limited exceptions. Active systems in passive ALWR designs provide the first line of defense to reduce challenges to the passive systems in the event of a transient at the nuclear power plant. Active systems that provide a defense-in-depth function in passive ALWR designs need not meet all of the acceptance criteria for safetyrelated systems. However, there should be a high level of confidence that these active systems will be available and reliable when challenged. The combined activities to provide confidence in the capability of these active systems in passive ALWR designs to perform their functions important to safety are referred to together as the RTNSS program. In a public memorandum dated July 24, 1995, the NRC staff provided a consolidated list of the approved policy and technical positions associated with RTNSS equipment in passive plant designs discussed in SECY–94–084 and SECY– 95–132 (ADAMS Accession No. ML003708048). This new paragraph will specify the need for licensees to assess the operational readiness of RTNSS pumps, valves, and dynamic restraints. 10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II) The NRC proposes to revise § 50.55a(b)(3)(iv) to address Appendix II, ‘‘Check Valve Condition Monitoring Program,’’ provided in the 2003 Addenda through the 2012 Edition of the ASME OM Code. In the 2003 Addenda of the ASME OM Code, ASME revised Appendix II to address the conditions specified in § 50.55a for older versions of the appendix. Therefore, the NRC considers Appendix E:\FR\FM\18SEP5.SGM 18SEP5 56832 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 II in the 2003 Addenda through the 2012 Edition of the ASME OM Code to be acceptable for use without conditions. In accepting the recent versions of Appendix II, the NRC proposes to clarify that (1) the maximum test interval allowed by Appendix II for individual check valves in a group of two valves or more must be supported by periodic testing of a sample of check valves in the group during the allowed interval and (2) the periodic testing plan must be designed to test each valve of a group at approximate equal intervals not to exceed the maximum requirement interval. The NRC notes that ASME has provided additional improvements to Appendix II since issuance of the 2003 Addenda. Therefore, where a licensee plans to voluntarily implement Appendix II to the ASME OM Code, the licensee should apply Appendix II in the most recent addenda and edition of ASME OM Code incorporated by reference in § 50.55a. The conditions currently specified for the use of Appendix II, 1995 Edition with the 1996 and 1997 Addenda, and 1998 Edition through the 2002 Addenda, of the OM Code remain the same in this proposed rule. 10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB The NRC proposes to add § 50.55a(b)(3)(vii) to prohibit the use of Subsection ISTB, ‘‘Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants,’’ in the 2011 Addenda of the ASME OM Code. In the 2011 Addenda to the ASME OM Code, the upper end of the Acceptable Range and the Required Action Range for flow and differential or discharge pressure for comprehensive pump testing in Subsection ISTB was raised to higher values. The NRC staff on the ASME OM Code committee accepted the proposed increase of the upper end of the Acceptable Range and Required Action Range with the planned addition of a requirement for a pump periodic verification test program in the ASME OM Code. However, the 2011 Addenda to the ASME OM Code did not include the requirement for a pump periodic verification test program as an oversight. Since then, the 2012 Edition to the ASME OM Code has incorporated Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program,’’ that supports the changes to the acceptable and required action ranges for comprehensive pump testing in Subsection ISTB. Therefore, proposed new § 50.55a(b)(3)(vii) would prohibit the use of Subsection ISTB in the 2011 Addenda of the ASME OM Code. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 Licensees will be allowed to apply Subsection ISTB with the revised acceptable and required action ranges in the 2012 Edition of the ASME OM Code as incorporated by reference in § 50.55a. 10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE The NRC proposes to add § 50.55a(b)(3)(viii) to specify that licensees proposing to implement Subsection ISTE, ‘‘Risk-Informed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,’’ of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition, must request and obtain NRC authorization in accordance with § 50.55a(z) to apply Subsection ISTE on a plant-specific basis as a risk-informed alternative to the applicable IST requirements in the ASME OM Code. In the 2009 Edition of the ASME OM Code, the ASME included new Subsection ISTE that describes a voluntary risk-informed approach in developing an IST program for pumps and valves at nuclear power plants. If a licensee chooses to implement this riskinformed IST approach, Subsection ISTE indicates that all requirements in Subsection ISTA, ‘‘General Requirements,’’ Subsection ISTB, and Subsection ISTC, ‘‘Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants,’’ of the ASME OM Code continue to apply, except those identified in Subsection ISTE. The ASME selected risk-informed guidance from ASME OM Code Cases OMN–1, OMN–3, OMN–4, ‘‘Requirements for Risk Insights for Inservice Testing of Check Valves at LWR Power Plants,’’ OMN–7, ‘‘Alternative Requirements for Pump Testing,’’ OMN–11, and OMN–12, ‘‘Alternative Requirements for Inservice Testing Using Risk Insights for Pneumatically and Hydraulically Operated Valve Assemblies in LightWater Reactor Power Plants,’’ in preparing Subsection ISTE of the ASME OM Code. During development of Subsection ISTE, the NRC staff participating on the ASME OM Code committees indicated that the conditions specified in RG 1.192 for the use of the applicable ASME OM Code Cases need to be considered when evaluating the acceptability of the implementation of Subsection ISTE. In addition, the NRC staff noted that several aspects of Subsection ISTE will need to be addressed on a case-by-case basis when determining the acceptability of its implementation. Therefore, new § 50.55a(b)(3)(viii) requires that licensees proposing to implement Subsection ISTE of the ASME OM Code PO 00000 Frm 00014 Fmt 4701 Sfmt 4702 must request approval from the NRC to apply Subsection ISTE on a plantspecific basis as a risk-informed alternative to the applicable IST requirements in the ASME OM Code. Nuclear power plant applicants for construction permits under 10 CFR part 50, or combined licenses for construction and operation under 10 CFR part 52, may describe their proposed implementation of the riskinformed IST approach specified in Subsection ISTE of the ASME OM Code for NRC review in their applications. The NRC will evaluate § 50.55a(z) requests for approval to implement Subsection ISTE in accordance with the following considerations: 1. Scope of Risk-Informed IST Program Subsection ISTE–1100, ‘‘Applicability,’’ establishes the component safety categorization methodology and process for dividing the population of pumps and valves, as identified in the IST Program Plan, into high safety significant component (HSSC) and low safety significant component (LSSC) categories. When establishing a risk-informed IST program, the licensee should address a wide range of components important to safety at the nuclear power plant that includes both safety-related and nonsafety-related components. These components might extend beyond the scope of the ASME OM Code. 2. Risk-Ranking Methodology The licensee should specify in its request for authorization to implement a risk-informed IST program the methodology to be applied in risk ranking its components. ISTE–4000, ‘‘Specific Component Categorization Requirements,’’ incorporates ASME OM Code Case OMN–3 for the categorization of pumps and valves in developing a risk-informed IST program. The OMN– 3 Code Case methodology for risk ranking uses two categories of safety significance. The NRC staff has also accepted other methodologies for risk ranking that use three categories of safety significance. 3. Safety Significance Categorization The licensee should categorize components according to their safety significance based on the methodology described in Subsection ISTE with the applicable conditions on the use of ASME OM Code Case OMN–3 specified in RG 1.192, or use other risk ranking methodologies accepted by the NRC on a plant-specific or industry-wide basis with applicable conditions specified by the NRC for their acceptance. The licensee should address the seven E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules conditions in RG 1.192 for the use of ASME OM Code Case OMN–3 as appropriate in developing the riskinformed IST program described in Subsection ISTE. With respect to the provisions in Subsection ISTE, these conditions are: (a) The implementation of ISTE–1100 should include within the scope of a licensee’s risk-informed IST program non-ASME Code pumps and valves categorized as HSSCs that might not currently be included in the IST program at the nuclear power plant. (b) The decision criteria discussed in ISTE–4410, ‘‘Decision Criteria,’’ and Non-mandatory Appendix L, ‘‘Acceptance Guidelines,’’ of the ASME OM Code for evaluating the acceptability of aggregate risk effects (i.e., for Core Damage Frequency [CDF] and Large Early Release Frequency [LERF]) should be consistent with the guidance provided in RG 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.’’ (c) The implementation of ISTE–4440, ‘‘Defense in Depth,’’ should be consistent with the guidance contained in Section 2.2.1, ‘‘Defense-in-Depth Evaluation,’’ and Section 2.2.2, ‘‘Safety Margin Evaluation,’’ of RG 1.175, ‘‘An Approach for Plant-Specific, RiskInformed Decisionmaking: Inservice Testing.’’ (d) The implementation of ISTE–4500, ‘‘Inservice Testing Program,’’ and ISTE– 6100, ‘‘Performance Monitoring,’’ should be consistent with the guidance contained in Section 3.2, ‘‘Program Implementation,’’ and Section 3.3, ‘‘Performance Monitoring,’’ of RG 1.175. (e) The implementation of ISTE–3210, ‘‘Plant-Specific PRA,’’ should be consistent with the guidance that the Owner is responsible for demonstrating and justifying the technical adequacy of the PRA analyses used as the basis to perform component risk ranking and for estimating the aggregate risk impact. For example, RG 1.200, ‘‘An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,’’ and RG 1.201, ‘‘Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance,’’ provide guidance for PRA technical adequacy and component risk ranking. (f) The implementation of ISTE–4240, ‘‘Reconciliation,’’ should specify that the expert panel may not classify components that are ranked HSSC by the results of a qualitative or quantitative PRA evaluation (excluding VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 the sensitivity studies) or the defensein-depth assessment to LSSC. (g) The implementation of ISTE–3220, ‘‘Living PRA,’’ should be consistent with the following: (i) To account for potential changes in failure rates and other changes that could affect the PRA, changes to the plant must be reviewed and, as appropriate, the PRA updated; (ii) when the PRA is updated, the categorization of structures, systems, and components must be reviewed and changed if necessary to remain consistent with the categorization process; and (iii) the review of the plant changes must be performed in a timely manner and must be performed once every two refueling outages, or as required by § 50.71(h)(2) for COL holders. 4. Pump Testing Subsection ISTE–5100, ‘‘Pumps,’’ incorporates ASME OM Code Case OMN–7 for risk-informed testing of pumps categorized as LSSCs. Subsection ISTE–5100 allows the interval for Group A and Group B testing of LSSC pumps specified in Subsection ISTB of the ASME OM Code to be extended from the current 3-month interval to intervals of 6 months or 2 years. Subsection ISTE–5100 eliminates the requirement in Subsection ISTB to perform comprehensive pump testing for LSSC pumps. Table ISTE–5121–1, ‘‘LSSC Pump Testing,’’ specifies that pump operation may be required more frequently than the specified test frequency (6 months) to meet vendor recommendations. Subsection ISTE– 4500, ‘‘Inservice Testing Program,’’ specifies in ISTE–4510, ‘‘Maximum Testing Interval,’’ that the maximum testing interval shall be based on the more limiting of (a) the results of the aggregate risk, or (b) the performance history of the component. ISTE–5130, ‘‘Maximum Test Interval—Pre-2000 Plants,’’ specifies that the most limiting interval for LSSC pump testing shall be determined from ISTE–4510 and ISTE– 5120, ‘‘Low Safety Significant Pump Testing.’’ The ASME developed the comprehensive pump test requirements in the ASME OM Code to address weaknesses in the Code requirements to assess the operational readiness of pumps to perform their design-basis safety function. Therefore, the licensee should ensure that testing under Subsection ISTE will provide assurance of the operational readiness of pumps in each safety significant categorization to perform their design-basis safety function as described in RGs 1.174 and 1.175. PO 00000 Frm 00015 Fmt 4701 Sfmt 4702 56833 5. Motor-Operated Valve Testing Subsection ISTE–5300, ‘‘Motor Operated Valve Assemblies,’’ provides a risk-informed IST approach instead of the IST requirements for MOVs in Mandatory Appendix III to the ASME OM Code. The ASME prepared Appendix III to the OM Code to replace the requirement for quarterly stroketime testing of MOVs with a program of periodic exercising and diagnostic testing to address lessons learned from nuclear power plant operating experience and industry and regulatory research programs for MOV performance. Subsection ISTC of the ASME OM Code specifies the implementation of Appendix III for periodic exercising and diagnostic testing of MOVs to replace quarterly stroke-time testing previously required for MOVs. Appendix III incorporates provisions that allow a risk-informed IST approach for MOVs as described in ASME OM Code Cases OMN–1 and OMN–11. Subsection ISTE–5300 is not consistent with the provisions for the risk-informed IST program for MOVs specified in Appendix III to the ASME OM Code (and Code Cases OMN–1 and 11). Therefore, licensees proposing to implement Subsection ISTE should address the provisions in paragraph III– 3700, ‘‘Risk-Informed MOV Inservice Testing,’’ of Appendix III to the ASME OM Code as incorporated by reference in § 50.55a with the applicable conditions instead of ISTE–5300. 6. Pneumatically and Hydraulically Operated Valve Testing Subsection ISTE–5400, ‘‘Pneumatically and Hydraulically Operated Valves,’’ specifies that licensees test their AOVs and HOVs in accordance with Appendix IV to the ASME OM Code. Subsection ISTE–5400 indicates that Appendix IV is in the course of preparation. The NRC staff will need to review Appendix IV prior to accepting its use as part of Subsection ISTE. Therefore, licensees proposing to implement Subsection ISTE should describe the planned IST provisions for AOVs and HOVs in its request for authorization to implement Subsection ISTE. 7. Pump Periodic Verification Test Subsection ISTE does not include a requirement to implement the pump periodic verification test program specified in Mandatory Appendix V to the ASME OM Code, 2012 Edition. The licensee should address the consideration of a pump periodic verification test program in its riskinformed IST program proposed as part E:\FR\FM\18SEP5.SGM 18SEP5 56834 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules of the authorization request to implement Subsection ISTE. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF The NRC proposes to add § 50.55a(b)(3)(ix) for two purposes. First, the proposed condition specifies that licensees applying Subsection ISTF, ‘‘Inservice Testing of Pumps in LightWater Reactor Nuclear Power Plants— Post-2000 Plants,’’ in the 2012 Edition of the OM Code shall satisfy the requirements of Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program,’’ of the OM Code, 2012 Edition. The proposed condition also states that Subsection ISTF, 2011 Addenda, is not acceptable for use. As previously discussed regarding new § 50.55a(b)(3)(vii), the upper end of the Acceptable Range and the Required Action Range for flow and differential or discharge pressure for comprehensive pump testing in Subsection ISTB in the ASME OM Code was raised to higher values in combination with the incorporation of Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program.’’ However, Subsection ISTF in the 2011 Addenda and 2012 Edition to the ASME OM Code does not include a requirement for a pump periodic verification test program. Therefore, new § 50.55a(b)(3)(ix) would require that the provisions of Appendix V be applied when implementing Subsection ISTF of the 2012 Edition of the OM Code to support the application of the upper end of the Acceptable Range and the Required Action Range for flow and differential or discharge pressure for inservice pump testing in Subsection ISTF. The proposed paragraph would prohibit the use of Subsection ISTF in the 2011 Addenda of the OM Code, which does not include Appendix V. 10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication The NRC proposes to add a new paragraph, § 50.55a(b)(3)(xi), containing a new condition that would specify that when implementing ASME OM Code, Subsection ISTC–3700, ‘‘Position Verification Testing,’’ licensees shall supplement the ASME OM Code provisions as necessary to verify that valve operation is accurately indicated. Subsection ISTC–3700 of the ASME OM Code requires that valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. Subsection ISTC–3700 states that where practicable, this local observation should be supplemented by other indications such as the use of flow meters or other suitable instrumentation VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 to verify obturator position. Subsection ISTC–3700 also states that where local observation is not possible, other indications shall be used for verification of valve operation. Nuclear power plant operating experience has revealed that reliance on indicating lights and stem travel are not sufficient to satisfy the requirement in ISTC–3700 to verify that valve operation is accurately indicated. Appendix A, ‘‘General Design Criteria for Nuclear Power Plants,’’ to 10 CFR part 50 requires that where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency, and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. This new condition specifies that when implementing ASME OM Code, Subsection ISTC–3700, licensees shall develop and implement a method to verify that valve operation is accurately indicated by supplementing valve position indicating lights with other indications, such as flow meters or other suitable instrumentation, to provide assurance of proper obturator position. This is not a new requirement but rather a clarification of the intent of the existing ASME OM Code. The ASME OM Code specifies obturator movement verification in order to detect certain internal valve failure modes consistent with the definition of ‘exercising’ found in ISTA–2000 (i.e., demonstration that the moving parts of a component function). Verification of the ability of an obturator to change or maintain position is an essential element of valve operational readiness determination which is a fundamental aspect of the ASME OM Code. The NRC’s position is further elaborated in NUREG–1482 Revision 2, paragraph 4.2.7. 10 CFR 50.55a(f): Inservice Testing Requirements The NRC proposes to revise the introductory text of § 50.55a(f) to indicate that systems and components must meet the requirements for ‘‘preservice and inservice testing’’ in the applicable ASME Codes and that both activities are referred to as ‘‘inservice testing’’ in the remainder of paragraph (f). The proposed change clarifies that the ASME OM Code includes provisions for preservice testing of components as part of its overall provisions for IST programs. No expansion of IST program scope is intended by this clarification. PO 00000 Frm 00016 Fmt 4701 Sfmt 4702 10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision The NRC proposes to revise § 50.55a(f)(3)(iii)(A) to ensure that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. Paragraph ISTA–1100, ‘‘Scope,’’ in Subsection ISTA, ‘‘General Requirements,’’ of the ASME OM Code states that the requirements for preservice and inservice testing and examination of components in lightwater reactor nuclear power plants apply to (a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident; (b) pressure relief devices that protect systems or portions of systems that perform one or more of these three functions; and (c) dynamic restraints (snubbers) used in systems that perform one of more of these three functions, or to ensure the integrity of the reactor coolant pressure boundary. This revision will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6, ‘‘Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints.’’ 10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision The NRC proposes to revise § 50.55a(f)(3)(iii)(B) to clarify that this paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This revision will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First Provision The NRC proposes to revise § 50.55a(f)(3)(iv)(A) to clarify that this paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(A) for Class 1 pumps and valves. This revision will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second Provision The NRC proposes to revise § 50.55a(f)(3)(iv)(B) to clarify that this paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(B) for Class 1 pumps E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules and valves. This revision will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for Operating Plants The NRC proposes to revise § 50.55a(f)(4) to clarify that this paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This revision will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 D. ASME Code Cases The NRC proposes to remove the revision number of the three RGs currently approved by the Office of the Federal Register for incorporation by reference throughout the substantive provisions of § 50.55a. The revision numbers for the RGs approved for incorporation by reference (currently, RGs 1.84, 1.147, and 1.192) would be retained in paragraph (a)(3)(i) through (a)(3)(iii) of § 50.55a, where the RGs are listed by full title, including revision number. These proposed changes would simplify the regulatory language containing cross-references to these RGs and reduce the possibility of NRC error in preparing future amendments to § 50.55a with respect to these RGs. These changes are administrative in nature and do not change substantive requirements with respect to the RGs and the Code Cases listed in the RGs. ASME BPV Code Case N–729–4 On September 10, 2008, the NRC issued a final rule to update § 50.55a to the 2004 Edition of the ASME Code (73 FR 52730). As part of the final rule, § 50.55a(g)(6)(ii)(D) implemented an augmented inservice inspection program for the examination of reactor pressure vessel (RPV) upper head penetration nozzles and associated partial penetration welds. The program required the implementation of ASME BPV Code Case N–729–1, with certain conditions. The application of ASME BPV Code Case N–729–1 was necessary because the inspections required by the 2004 Edition of the ASME BPV Code, Section XI were not written to address degradation of the RPV upper head penetration nozzles welds by primary water stress corrosion cracking (PWSCC). The safety consequences of inadequate inspections can be significant. The NRC’s determination that the ASME Code required VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 inspections are inadequate is based upon operating experience and analysis. The absence of an effective inspection regime could, over time, result in unacceptable circumferential cracking, or the degradation of the RPV upper head or other reactor coolant system components by leakage assisted corrosion. These degradation mechanisms increase the probability of a loss-of-coolant accident. Examination frequencies and methods for RPV upper head penetration nozzles and welds are provided in ASME BPV Code Case N–729–1. The use of code cases is voluntary, so these provisions were developed, in part, with the expectation that the NRC would incorporate the code case by reference into the CFR. Therefore, the NRC adopted rule language in § 50.55a(g)(6)(ii)(D) requiring implementation of ASME BPV Code Case N–729–1, with conditions, in order to enhance the examination requirements in the ASME BPV Code, Section XI for RPV upper head penetration nozzles and welds. The examinations conducted in accordance with ASME BPV Code Case N–729–1 provide reasonable assurance that ASME Code allowable limits will not be exceeded and that PWSCC will not lead to failure of the RPV upper head penetration nozzles or welds. However, the NRC concluded that certain conditions were needed in implementing the examinations in ASME BPV Code Case N–729–1. These conditions are set forth in § 50.55a(g)(6)(ii)(D). On June 22, 2012, the ASME approved the fourth revision of ASME BPV Code Case N–729, (N–729–4). This revision changed certain requirements based on a consensus review of inspection techniques and frequencies. These changes were deemed necessary by the ASME to supersede the previous requirements under N–729–1 to establish an effective long-term inspection program for the RPV upper head penetration nozzles and associated welds in pressurized water reactors. The major changes included incorporation of previous NRC conditions in the CFR. Minor changes were also made to address editorial issues, to correct figures or to add clarity. The NRC proposes to update the requirements of § 50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV Code Case N–729–4, with conditions. The NRC conditions have been modified to address the changes in ASME BPV Code Case N–729–4. The NRC’s proposed revisions to the conditions on ASME BPV Code Case N– PO 00000 Frm 00017 Fmt 4701 Sfmt 4702 56835 729–1 are discussed in the next four sections. 10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(1) to change the version of ASME BPV Code Case N–729 from N–729–1 to N–729–4 for the reasons previously set forth. Due to the incorporation of N–729–4, the date to establish applicability for licensed pressurized water reactors will be changed to the effective date of the final rule. 10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6) The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(2) through (6) to remove the conditions currently in § 50.55a(g)(6)(ii)(D)(2) through (5) and to redesignate the condition currently in § 50.55a(g)(6)(ii)(D)(6) as § 50.55a(g)(6)(ii)(D)(2). The conditions currently in § 50.55a(g)(6)(ii)(D)(2) to § 50.55a(g)(6)(ii)(D)(5) have all been incorporated either verbatim or more conservatively in the revisions to ASME BPV Code Case N–729, up to version N– 729–4. Therefore, there is no reason to retain these conditions in § 50.55a. The NRC proposes to include new conditions in § 50.55a(g)(6)(ii)(D)(3) and (4) as described in the following discussion. 10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency The NRC proposes to adopt a new condition (to be included in proposed § 50.55a(g)(6)(ii)(D)(3)) to modify the option to extend bare metal visual inspections of the reactor pressure vessel upper head surface beyond the frequency listed in Table 1 of ASME BPV Code Case N–729–4. Previously, upper heads aged with less than eight effective degradation years were considered to have a low probability of initiating PWSCC, the cracking mechanism of concern. This ranking of effective degradation years was based on a simple time at temperature correlation. All of the upper heads within this category, with the exception of new heads using Alloy 600 penetration nozzles, were considered to have lower susceptibility to cracking due to the upper heads being at or near the cold leg operating temperature of the reactor coolant system. Therefore, these plants were said to have ‘‘cold heads.’’ All of the upper heads that had experienced cracking prior to 2006 were near the hot leg operating temperature of the reactor coolant system, which validated the time at temperature model. E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56836 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules In 2006, one of the 21 ‘‘cold head’’ plants identified two indications within a penetration nozzle and the associated partial penetration weld. Then, between 2006 and 2013, five of the 21 ‘‘cold head’’ plants identified multiple indications within fifteen different penetration nozzles and the associated partial penetration welds. None of these indications caused leakage, and volumetric examination of the penetration nozzles showed no flaw in the nozzle material had grown throughwall; however, this increasing trend creates a reasonable safety concern. Recent operational experience has shown that the volumetric inspection of penetration nozzles, at the current inspection frequency, is adequate to identify indications in the nozzle material prior to leakage; however, volumetric examinations cannot be performed on the partial penetration welds. Therefore, given the additional cracking identified at cold leg temperature, the NRC staff has concerns about the adequacy of the partial penetration weld examinations. Leakage from a partial penetration weld into the annulus between the nozzle and head material can cause corrosion of the low alloy steel head. While initially limited in leak rate, due to limited surface area of the weld being in contact with the annulus region, corrosion of the vessel head material can expose more of the weld surface to the annulus, allowing a greater leak rate. Since an indication in the weld cannot be identified by a volumetric inspection, a postulated crack through the weld, just about to cause leakage, could exist as a plant performed its last volumetric and/or bare metal visual examination of the upper head material. This gives the crack years to breach the surface and leak prior to the next scheduled visual examination. Only a surface examination of the wetted surface of the partial penetration weld can reliably detect flaws in the weld. Unfortunately, this examination cannot size the flaws in the weld, and, if performed manually, requires significant radiological dose to examine all the partial penetration welds on the upper head. As such, the available techniques are only able to detect a flaw after it has caused leakage. These techniques are a bare metal visual examination or a volumetric leak path assessment performed on the frequency of the volumetric examination. Volumetric leak path examinations are only done on outages when a volumetric examination of the nozzle is performed. Therefore, under the current requirements allowed by Note 4 of ASME BPV Code Case N–729–4, leakage VerDate Sep<11>2014 22:31 Sep 17, 2015 Jkt 235001 from a crack in the weld of a ‘‘cold head’’ plant could start and continue to grow for the 5 years between the required bare metal visual examinations to detect leakage through the partial penetration weld. Given the additional cracking identified at cold leg temperature of upper head penetration nozzles and associated welds, the NRC finds limited basis to continue to categorize these ‘‘cold head’’ plants as having a low susceptibility to crack initiation. The NRC proposes to increase the frequency of the bare metal visual examinations of ‘‘cold heads’’ to identify potential leakage as soon as reasonably possible because of the volumetric examination limitations. Therefore, the NRC proposes to condition Note 4 of ASME BPV Code Case N–729–4 to require a bare metal visual exam each outage in which a volumetric exam is not performed. The NRC also proposes to allow ‘‘cold head’’ plants to extend their bare metal visual inspection frequency from once each refueling outage, as stated in Table 1 of N–729–1, to once every 5 years, but only if the licensee performed a wetted surface examination of all of the partial penetration welds during the previous volumetric examination. Applying the conditioned bare metal visual inspection frequency or a volumetric examination each outage will allow licensees to identify any potential leakage through the partial penetration welds prior to significant degradation of the low alloy steel head material, thereby providing reasonable assurance of the structural integrity of the reactor coolant pressure boundary. These issues, including the operational experience, the fact that volumetric examination is not available to interrogate the partial penetration welds, and potential regulatory options, were discussed publicly at multiple ASME Code meetings, at the annual Materials Programs Technical Information Exchange public meeting held at the NRC Headquarters in June 2013, and at the 2013 NRC Regulatory Information Conference. 10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria The NRC proposes to adopt a new condition (to be included in proposed § 50.55a(g)(6)(ii)(D)(4)) to define surface examination acceptance criteria. Paragraph –3132(b) of ASME BPV Code Case N–729–4 sets forth the acceptance criteria for surface examinations. In general, throughout Section XI of the ASME BPV Code, the acceptance criteria for surface examinations default to Section III, Paragraph NB–5352, ‘‘Acceptance Standards’’. Typically, for PO 00000 Frm 00018 Fmt 4701 Sfmt 4702 rounded indications, the indication was only unacceptable if it was greater than 3⁄16 inch in size. The NRC requested that the code case authors include a requirement that any size rounded indication causing nozzle leakage is unacceptable due to operating experience identifying PWSCC under rounded indications less than 3⁄16 inch in size. Recently, the ASME Code Committee approved an interpretation of the language in Paragraph –3132(b) that implied any size rounded indication is acceptable unless there is relevant indication of nozzle leakage, even those greater than 3⁄16 inch. The NRC does not agree with the interpretation and maintains its original stance on rounded indications that any size rounded indication is unacceptable if there is an indication of leakage. Since the adoption of ASME BPV Code Case N– 729–1 into § 50.55a(g)(6)(ii)(D), all licensees have used the NRC’s stance in implementing Paragraph –3132(b), even after the recent ASME Code Committee interpretation approval over NRC objection. Therefore, in order to ensure compliance with the previous and ongoing requirement, the NRC proposes to revise condition § 50.55a(g)(6)(ii)(D)(4) to include clarity within the acceptance criteria for surface examinations. The current edition requirements of NB–5352 of ASME BPV Code, Section III for the licensee’s ongoing 10-year inservice inspection interval shall be met. ASME BPV Code Case N–770–2 On June 21, 2011, the NRC issued a final rule including § 50.55a(g)(6)(ii)(F) requiring the implementation of ASME BPV Code Case N–770–1, ‘‘Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS N86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,’’ with certain conditions. On June 9, 2011, the ASME approved the second revision of ASME BPV Code Case N–770 (N–770–2). The major changes from N–770–1 to N–770–2 included establishing new ASME Code Case Table 1 inspection item classifications for optimized weld overlays and allowing alternatives when complete inspection coverage cannot be met. Minor changes were also made to address editorial issues, to correct figures, or to add clarity. The NRC finds that the updates and improvements in N–770–2 are sufficient to update § 50.55a(g)(6)(ii)(F). E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules The NRC therefore proposes to update the requirements of § 50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code Case N–770–2 with conditions. The NRC conditions have been modified to address the changes in ASME BPV Code Case N–770–2 and to ensure that this regulatory framework will provide adequate protection of public health and safety. The following sections discuss each of the NRC’s proposed changes to the conditions on ASME BPV Code Case N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(1) to change the version of ASME BPV Code Case N–770 from N–770–1 to N–770–2 and to require its implementation (with conditions) to incorporate the updates and improvements contained in N–770– 2. The NRC proposes that licensees begin using N–770–2 on the effective date of this rule. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(2) to provide clarification regarding categorization of each Alloy 82/182 butt weld, mitigated or not, under N–770–2. This paragraph also clarifies the NRC’s position that paragraph –1100(e) shall not be used to exempt welds that rely on Alloy 82/182 for structural integrity from more frequent ISI schedules until the NRC has reviewed and authorized an alternative categorization for the weld. Additionally, the NRC proposes to change the inspection item categories for full structural weld overlays from C to C–1 and F to F–1 due to reclassification under ASME BPV Code Case N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(3) to clarify the baseline examination requirements by stating that previously-conducted examinations, in order to count as baseline examinations, must meet the requirements of ASME BPV Code Case N–770–2, as conditioned. The 2011 rule required the use of ASME Code Section XI Appendix VIII qualifications for baseline examinations, which is stricter than N–770–2 and does not provide requirements for optimized weld overlays. The revision also updates the deadline for baseline examination requirements since the January 20, 2012, deadline from the previous rule has passed. Finally, upon implementation of this rule, if a licensee is currently in an VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 outage, then the baseline inspection requirement can be met by performing the inspections in accordance with the current regulatory requirements of § 50.55a(g)(6)(ii)(F) in lieu of the examination requirements of paragraphs –2500(a) or –2500(b) of ASME BPV Code Case N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(4) to define examination coverage for circumferential flaws and to prohibit the use of paragraph –2500(d) of ASME BPV Code Case N–770–2 which, in some circumstances, allows unacceptably low examination coverage. Paragraph –2500(d) of N–770–2 would allow the reduction of circumferential volumetric examination coverage with analytical evaluation. Paragraph –2500(c) was previously prohibited from use, and it continues to be prohibited. The NRC proposes to establish an essentially 100 percent volumetric examination coverage requirement for circumferential flaws to provide reasonable assurance of structural integrity of all ASME Code Class 1 butt welds susceptible to PWSCC. Therefore, the NRC proposes to adopt a condition prohibiting the use of paragraphs –2500(c) and –2500(d). A licensee may request approval for use of these paragraphs under 10 CFR 50.55a(z). 10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(5) to add the explanatory heading, ‘‘Inlay/onlay inspection frequency,’’ and to make minor editorial corrections. 10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(6) to add the explanatory heading, ‘‘Reporting requirements.’’ 10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ‘‘t’’ The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(7) to add the explanatory heading, ‘‘Defining ‘t’.’’ 10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(8) to maintain the requirement for the timing of the initial inservice examination of optimized weld overlays. Uncracked welds mitigated with optimized weld overlays were re-categorized by ASME BPV Code Case N–770–2 from Inspection Item D to PO 00000 Frm 00019 Fmt 4701 Sfmt 4702 56837 Inspection Item C–2; however, the initial inspection requirement was not incorporated into the Code Case for Inspection Item C–2. The NRC has determined that uncracked welds mitigated with an optimized weld overlay must have an initial inservice examination no sooner than the third refueling outage and no later than 10 years following the application of the weld overlay to identify unacceptable crack growth. Optimized weld overlays establish compressive stress on the inner half thickness of the weld, but the outer half thickness may also be under tensile stresses. The requirement for an initial inservice examination no sooner than the third refueling outage and no later than 10 years following the application of the weld overlay is based on the design of optimized weld overlays which require the outer quarter thickness of the susceptible material to provide structural integrity for the weld. Therefore, the NRC proposes to continue adoption of the condition which requires the initial inservice examination of uncracked welds mitigated by optimized weld overlay (i.e., the welds which are subject to Inspection Item C–2 of ASME BPV Code Case N–770–2) within the specified timeframe. 10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(9) to address changes in ASME BPV Code Case N– 770–2 which allow the deferral of the first inservice examination of uncracked welds mitigated with optimized weld overlays, Inspection Item C–2. Previously, under N–770–1, the initial inservice examination of these welds was not allowed to be deferred. Allowing deferral of the initial inservice examination in accordance with N–770– 2 could, in certain circumstances, allow the initial inservice examination to be performed up to 20 years after installation. Therefore, the NRC proposes to adopt a condition which would preclude the deferral of the initial inservice examination of uncracked welds mitigated by optimized weld overlays. 10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(10) to address changes in ASME BPV Code Case N– 770–2. Note 14(a) of Table 1 of ASME BPV Code Case N–770–2 provides the previously required full examination requirement for optimized weld overlays. The language of ASME BPV Code Case N–770–2, however, does not E:\FR\FM\18SEP5.SGM 18SEP5 56838 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules require the implementation of the full examination requirements of Note 14(a) of Table 1, if possible, before implementing the reduced examination coverage requirements of Note 14(b) of Table 1 or Note (b) of Figure 5(a). The full examination requirement should be implemented, if possible, before the option of reduced examination coverage is allowed. Therefore, the NRC proposes to modify the current condition in § 50.55a(g)(6)(ii)(F)(10) to allow the use of Note 14(b) of Table 1 and Note (b) of Figure 5(a) of ASME BPV Code Case N– 770–2 only after the determination that the requirements of Note 14(a) of Table 1 of ASME BPV Code Case N–770–2 cannot be met. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel The NRC proposes to add § 50.55a(g)(6)(ii)(F)(11) to address examination requirements through cast stainless steel materials by requiring the use of Appendix VIII qualifications to meet the inspection requirements of paragraph –2500(a) of ASME BPV Code Case N–770–2. The requirements for volumetric examination of butt welds through cast stainless steel materials are currently being developed as Supplement 9 to the ASME BPV Code, Section XI, Appendix VIII. In accordance with Appendix VIII for supplements that have not been developed, the requirements of Appendix III apply. Appendix III requirements are not equivalent to Appendix VIII requirements. For the volumetric examination of ASME Class 1 welds, the NRC has established the requirement for examination qualification under the Appendix VIII. Thus, the NRC proposes to adopt a condition requiring the use of Appendix VIII qualifications to meet the inspection requirements of paragraph –2500(a) of ASME BPV Code Case N– 770–2 by January 1, 2020. The development of a sufficient number of mockups would be required to establish an Appendix VIII program for examination of ASME Code Class 1 piping and vessel nozzle butt welds through cast stainless steel materials. The NRC recognizes that significant time and resources are required to create mockups and to allow for qualification of equipment, procedures and personnel. Therefore, the NRC proposes that licensees be required to use these Appendix VIII qualifications no later than their first scheduled weld examinations involving cast stainless steel materials occurring after January 1, 2020. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage The NRC proposes to add § 50.55a(g)(6)(ii)(F)(12) to clarify the examination coverage requirements allowed under Appendix I of ASME BPV Code Case N–770–2 for butt welds joining cast stainless steel material. Under current ASME BPV Code, Section XI, Appendix VIII requirements, the volumetric examination of butt welds through cast stainless steel materials is under Supplement 9. Supplement 9 rules are still being developed by the ASME BPV Code. Therefore, it is currently impossible to meet the requirement of Paragraph I.5.1 for butt welds joining cast stainless steel material. The material of concern is the weld material susceptible to PWSCC adjoining the cast stainless steel material. Appendix VIII qualified procedures are available to perform the inspection of the susceptible weld material, but they are not qualified to inspect the cast stainless steel materials. Therefore, the NRC proposes to adopt a condition changing the inspection volume for stress-improved dissimilar metal welds with cast stainless steel from the ASME Code Section XI requirements to ‘‘the maximum extent practical including 100 percent of the susceptible material volume.’’ This will remain applicable until an Appendix VIII qualified procedure for the inspection through cast stainless steel materials is available in accordance with the proposed condition in § 50.55a(g)(6)(ii)(F)(11). 10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination The NRC proposes to add § 50.55a(g)(6)(ii)(F)(13) to require the encoding of ultrasonic volumetric examinations of Inspection Items A–1, A–2, B, E, F–2, J, and K in Table 1 of N–770–2. A human performance gap has been found between some ultrasonic testing procedures as demonstrated during ASME BPV Code, Section XI, Appendix VIII qualification versus as applied in the field. The human factors that contributed to the recent examinations that failed to identify significant flaws at North Anna Power Station, Unit 1, in 2012 (Licensee Event Report 50–338/2012–001–00, ADAMS Accession No. ML12151A441) and at Diablo Canyon Nuclear Power Plant in 2013 (Relief Request REP–1 U2, Revision 2, ADAMS Accession No. ML13232A308) can be avoided by the use of encoded ultrasonic examinations. Encoded ultrasonic examinations electronically store both the positional PO 00000 Frm 00020 Fmt 4701 Sfmt 4702 and ultrasonic information from the inspections. Encoded examinations allow for the inspector to evaluate the data and search for indications outside of a time limiting environment to assure that the inspection was conducted properly and to allow for sufficient time to analyze the data. Additionally, the encoded examination would allow for an independent review of the data by other inspectors or an independent third party. Finally, the encoded examination could be compared to previous and/or future encoded examinations to determine if flaws are present and flaw growth rate. Therefore, the NRC proposes to adopt a condition requiring the use of encoding for ultrasonic volumetric examinations of nonmitigated or cracked mitigated dissimilar metal butt welds in the reactor coolant pressure boundary which are within the scope of ASME BPV Code Case N–770–2. ASME BPV Code Case N–824 10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N–824 The NRC proposes to add § 50.55a(b)(2)(xxxvii) to allow licensees to use the provisions of ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ subject to NRCproposed conditions of § 50.55a(b)(2)(xxxvii)(A) through (E), when implementing inservice examinations in accordance with the ASME BPV Code, Section XI requirements. During the construction of nuclear power plants, it was recognized that the grain structure of cast austenitic stainless steel (CASS) could prevent effective ultrasonic inspections of piping welds where one or both sides of the welds were constructed of CASS. The high strength and toughness of CASS (prior to thermal embrittlement) made it desirable as a building material despite this known inspection issue. This choice of construction materials has rendered many pressure boundary components without a means to reliably inspect them volumetrically. While there is no operational experience of a CASS component failing, as part of the reactor pressure boundary, inservice volumetric inspection of these components is necessary to provide reasonable assurance of their structural integrity. The current regulatory requirements for the examination of CASS, provided in § 50.55a, do not provide sufficient guidance to assure that the CASS components are being inspected E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules adequately. To illustrate that ASME Code does not provide adequate guidance, ASME Code, Section XI, Appendix III, Supplement 1 states ‘‘Cast materials may preclude meaningful examinations because of geometry and attenuation variables.’’ For this reason, over the past several decades, licensees have been unable to perform effective inspections of welds joining CASS components. To allow for continued operation of their plants, licensees submitted hundreds of requests for relief from the ASME Code requirements for inservice inspection of CASS components to the NRC, resulting in a significant regulatory burden. Based on the improvements in ultrasonic inspection technology and techniques for CASS components, the ASME approved BPV Code Case N–824 (N– 824) on October 16, 2012, which describes how to develop a procedure capable of meaningfully inspecting welds in CASS components. The NRC commissioned a research program to determine the effectiveness of the new technologies for inspections of CASS components in an effort to resolve some of the known inspection issues. The result of this work is published in NUREG/CR–6933, ‘‘Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced LowFrequency Ultrasonic Methods’’, March 2007, and NUREG/CR–7122, ‘‘An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,’’ March 2012. These NUREG/CR reports show that CASS materials less than 1.6 inches thick can be reliably inspected for flaws 10 percent throughwall or deeper if encoded phased-array examinations are performed using low ultrasonic frequencies and a sufficient number of inspection angles. Additionally, for thicker welds, flaws greater than 30 percent through-wall in depth can be detected using low frequency encoded phased-array ultrasonic inspections. The NRC, using NUREG/CR–6933 and NUREG/CR–7122, has determined that inspections of CASS materials are very challenging, and sufficient technical basis exists to condition the code case to bring the code case into agreement with the NUREG/CR reports. The NUREG/CR reports also show that CASS materials produce high levels of coherent noise. The noise signals can be confusing and mask flaw indications. Use of encoded inspection data allows the inspector to mitigate this problem through the ability to electronically manipulate the data, which allows for discrimination between coherent noise VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 and flaw indications. The NRC finds that encoding CASS inspection data provides significant detection benefits. The NRC proposes to add a condition in § 50.55a(b)(2)(xxxvii)(A) to require the use of encoded data when utilizing N– 824 for the examination of CASS components. The use of dual element phased-array search units showed the most promise in obtaining meaningful responses from flaws. The NRC proposes to add a condition in § 50.55a(b)(2)(xxxvii)(B) to require the use of dual, transmit-receive, refracted longitudinal wave, multi-element phased array search units when utilizing N–824 for the examination of CASS components. The optimum inspection frequencies for examining CASS components of various thicknesses as described in NUREG/CR– 6933 and NUREG/CR–7122 are reflected in proposed conditions § 50.55a(b)(2)(xxxvii)(C) and (D). The NRC proposes to add a condition in § 50.55a(b)(2)(xxxvii)(C) to require that ultrasonic examinations performed to implement ASME BPV Code Case N– 824 on piping less than or equal to 1.6 inches thick shall use a phased array search unit with a center frequency of 500 kHz to 1 MHz. The NRC proposes to add a condition in § 50.55a(b)(2)(xxxvii)(D) to require that ultrasonic examinations performed to implement ASME BPV Code Case N– 824 on piping greater than 1.6 inches thick shall use a phased array search unit with a center frequency of 500 kHz. As NUREG/CR–6933 shows that the grain structure of CASS can reduce the effectiveness of some inspection angles, the NRC finds sufficient technical basis to condition the code case for the use of phased-array ultrasound using angles from 30 to 70 degrees with a maximum increment of 5 degrees. The NRC proposes to add a condition in § 50.55a(b)(2)(xxxvii)(E) to require that ultrasonic examinations performed to implement ASME BPV Code Case N– 824 shall use a phased array search unit which produces angles from 30 to 70 degrees with a maximum increment of 5 degrees. Obtaining effective examination results of CASS components requires using lower frequencies and larger transducers than are typically used for ultrasonic inspections of piping welds and would require licensees to modify their inspection procedures. The NRC recognizes that requiring the use of spatial encoding will limit the full implementation of ASME BPV Code Case N–824, as spatial encoding is not practical for many weld configurations. The recent advances in inspection technology are driving renewed work at PO 00000 Frm 00021 Fmt 4701 Sfmt 4702 56839 ASME Code meetings to produce Section XI, Appendix VIII, Supplement 9 to resolve the CASS inspection issue, but it will be years before these code updates will be published, as well as additional time to qualify and approve procedures for use in the field. Until then, licensees would still use the requirements of ASME Code Section XI, Appendix III, Supplement 1 which states that inspection of CASS materials meeting the ASME Code requirements may not be meaningful. Consequently, less effective examinations would continue to be used in the field, and more relief requests would be generated between now and the implementation of Supplement 9. At this time, the use of ASME BPV Code Case N–824, as conditioned, is the most effective known method for adequately examining welds with one or more CASS components. With the use of ASME BPV Code Case N–824, as conditioned, licensees will be able to take full credit for completion of the § 50.55a required inservice volumetric inspection of welds involving CASS components. The implementation of ASME BPV Code Case N–824, as conditioned, will have the dual effect of improving the rigor of required volumetric inspections and reducing the number of uninspectable Class 1 and Class 2 pressure retaining welds. The NRC concludes that incorporation of ASME BPV Code Case N–824, as conditioned by § 50.55a(b)(2)(xxxvii)(A) through (E), will significantly improve the flaw detection capability of ultrasonic inspection of CASS components until Supplement 9 is implemented, thereby providing reasonable assurance of leak tightness and structural integrity. Additionally, it will reduce the regulatory burden on licensees and allow licensees to submit fewer relief requests for welds in CASS materials. ASME OM Code Case OMN–20 10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN–20 The NRC proposes to add new paragraph § 50.55a(b)(3)(x) to allow the use of ASME OM Code Case OMN–20, ‘‘Inservice Test Frequency,’’ which provides inservice test frequencies for pumps and valves which a licensee may voluntarily use in place of the frequencies specified in the 2012 Edition of the ASME OM Code. Paragraph § 50.55a(a)(1)(iii)(E) would be added to incorporate ASME OM Code Case OMN–20 by reference into § 50.55a. Surveillance Requirement (SR) 3.0.3 from Technical Specification (TS) 5.5.6, ‘‘Inservice Testing Program,’’ E:\FR\FM\18SEP5.SGM 18SEP5 56840 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules allows licensees to apply a delay period before declaring the SR for TS equipment ‘‘not met’’ when the licensee inadvertently exceeds or misses the time limit for performing TS surveillance. Licensees have been applying SR 3.0.3 to inservice tests. The NRC has determined that licensees cannot use TS 5.5.6 to apply SR 3.0.3 to inservice tests under § 50.55a(f) that are not associated with a TS surveillance. To invoke SR 3.0.3, the licensee shall first discover that a TS surveillance was not performed at its specified frequency. Therefore, the delay period that SR 3.0.3 provides does not apply to non-TS support components tested under § 50.55a(f). The ASME OM Code does not provide for any inservice test frequency reductions or extensions. In order to provide inservice test frequency reductions or extensions that can no longer be provided by SR 3.0.3 from TS 5.5.6, the ASME has developed OM Code Case OMN–20. The NRC has reviewed OM Code Case OMN–20 and has found it acceptable for use. The NRC intends to include OM Code Case OMN–20 in the next revision of RG 1.192, at which time a conforming change will be made to delete both this paragraph and § 50.55a(a)(1)(iii)(E). mstockstill on DSK4VPTVN1PROD with PROPOSALS5 IV. Section-by-Section Analysis The NRC proposes to remove the revision number of the three RGs currently approved by the Office of the Federal Register for incorporation by reference throughout the substantive provisions of § 50.55a. The revision numbers for the RGs approved for incorporation by reference would be retained in paragraph (a) of § 50.55a, where the RGs are listed by full title, including revision number. That paragraph identifies the specific materials which the Office of the Federal Register has approved for incorporation by reference, as required by Office of the Federal Register requirements in 1 CFR 51.9. No substantive change is intended by the NRC by this proposed amendment. Readers would need to refer to paragraph (a) of § 50.55a to determine the specific revision of the relevant RG which is approved for incorporation by reference by Office of the Federal Register. 10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section III The NRC proposes to revise § 50.55a(a)(1)(i) to clarify that Section III Nonmandatory Appendices are not incorporated by reference. This language was originally added in a final rule published on June 21, 2011 (76 FR 36232); however, it was omitted from the final rule published on November 5, 2014 (79 FR 65776). The NRC is correcting the omission by inserting ‘‘(excluding Nonmandatory Appendices)’’ in 10 CFR 50.55a(a)(1)(i). ‘‘ASME OM Code Case OMN–20,’’ and include information for the standard that is being incorporated by reference. 10 CFR 50.55a(a)(1)(i)(E) ‘‘Rules for Construction of Nuclear Facility Components—Division 1’’ The NRC proposes to revise § 50.55a(a)(1)(i)(E) to add ASME BPV Code, Section III 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition. The NRC proposes to revise § 50.55a(a)(1)(iv)(B) to add ASME OM Code 2009 Edition and 2011 Addenda. 10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section XI The NRC proposes to revise § 50.55a(a)(1)(ii) to include a minor editorial change and to clarify that Nonmandatory Appendix U is not incorporated by reference. 10 CFR 50.55a(a)(1)(ii)(C) ‘‘Rules for Inservice Inspection of Nuclear Power Plant Components—Division 1’’ The NRC proposes to revise § 50.55a(a)(1)(ii)(C) to add ASME BPV Code, Section XI 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition. 10 CFR 50.55a(a)(1)(iii)(B) ASME BPV Code Case N–729–4 The NRC proposes to revise § 50.55a(a)(1)(iii)(B) to add the title ‘‘ASME BPV Code Case N–729–4,’’ and include information for the standard that is being incorporated by reference. 10 CFR 50.55a(a)(1)(iii)(C) ASME BPV Code Case N–770–2 The NRC proposes to revise § 50.55a(a)(1)(iii)(C) to add the title ‘‘ASME BPV Code Case N–770–2,’’ and include information for the standard that is being incorporated by reference. 10 CFR 50.55a(a) Documents Approved for Incorporation by Reference 10 CFR 50.55a(a)(1)(iii)(D) ASME BPV Code Case N–824 The NRC proposes to add § 50.55a(a)(1)(iii)(D) to add the title ‘‘ASME BPV Code Case N–824,’’ and include information for the standard that is being incorporated by reference. The NRC proposes to revise the incorporation by reference language to update the contact information for the NRC Technical Library. 10 CFR 50.55a(a)(1)(iii)(E) ASME OM Code Case OMN–20 The NRC proposes to add § 50.55a(a)(1)(iii)(E) to add the title VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 PO 00000 Frm 00022 Fmt 4701 Sfmt 4702 10 CFR 50.55a(a)(1)(iv) ASME Operation and Maintenance Code The NRC proposes to revise § 50.55a(a)(1)(iv) to correct the title of the OM Code. 10 CFR 50.55a(a)(1)(iv)(B) ‘‘Operation and Maintenance of Nuclear Power Plants, Division 1: Section IST Rules for Inservice Testing of Light-Water Reactor Power Plants’’ 10 CFR 50.55a(a)(1)(iv)(C) ‘‘Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST’’ The NRC proposes to add § 50.55a(a)(1)(iv)(C) to add ASME OM Code 2012 Edition. 10 CFR 50.55a(a)(1)(v) ASME Quality Assurance Requirements The NRC proposes to add § 50.55a(a)(1)(v) to add the title ‘‘ASME Quality Assurance Requirements’’ for ASME NQA–1 Code as part of NRC titling convention and include information regarding NQA–1 standards. 10 CFR 50.55a(b) Use and Conditions on the Use of Standards The NRC proposes to revise § 50.55a(b) to correct the title of the OM Code. 10 CFR 50.55a(b)(1) Conditions on ASME BPV Code Section III The NRC proposes to revise § 50.55a(b)(1) to reflect the latest edition incorporated by reference, the 2013 Edition. 10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions The NRC proposes to revise § 50.55a(b)(1)(ii) to clarify rule language and add Table 1, which clarifies prohibited Section III provisions in tabular form for welds with leg size less than 1.09 tn. 10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance The NRC proposes to revise § 50.55a(b)(1)(iv) to clarify that it allows, but does not require, applicants and licensees to use the 2008 Edition through the 2009–1a Addenda of NQA– 1 when applying the 2010 Edition and later editions of the ASME BPV Code, Section III, up to the 2011 Addenda. E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules Applicants and licensees are required to meet appendix B of 10 CFR part 50, and NQA–1 is one way of meeting portions of appendix B. An applicant or licensee may select any version of NQA–1 that has been approved for use in § 50.55a, but they must also use the administrative, quality, and technical provisions contained in the version of NCA–4000 referencing that Edition or Addenda of NQA–1 selected by the applicant or licensee. NQA–1 provides a method for establishing and implementing a QA program for the design and construction of nuclear power plants and fuel reprocessing plants; however, NQA–1, as modified and supplemented by NCA– 4000, does not meet all of the requirements of appendix B to 10 CFR part 50. To meet the requirements of appendix B, when using NQA–1 during the design and construction phase, applicants and licensees must address in their quality program description those areas where NQA–1 is insufficient to meet appendix B. Regulatory Guide 1.28, ‘‘Quality Assurance Criteria (Design and Construction),’’ provides additional guidance and regulatory positions on how to meet appendix B when using NQA–1. Section 50.55a(b)(1)(iv) clarifies that applicants and licensees are required to meet appendix B to 10 CFR part 50 and that the commitments contained in their QA program descriptions that are more stringent than those contained in NQA– 1 or are not addressed in NQA–1 apply to Section III activities. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification and Demonstration of Function of Incompressible-Fluid Pressure-Relief Valves The NRC proposes to revise § 50.55a(b)(1)(vii) to reflect the latest edition incorporated by reference, the 2013 Edition. 10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME Certification Marks The NRC proposes to add § 50.55a(b)(1)(viii) to allow licensees to use either the ASME BPV Code Symbol Stamp or ASME Certification Mark with the appropriate certification designator and class designator as specified in the 2013 Edition through the latest edition and addenda incorporated by reference in 10 CFR 50.55a. 10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI The NRC proposes to revise § 50.55a(b)(2) to reflect the latest edition incorporated by reference, the 2013 VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 Edition, and to clarify that Nonmandatory Appendix U is not incorporated by reference. 10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and Addenda of Subsection IWE and Subsection IWL The NRC proposes to revise § 50.55a(b)(2)(vi) to clarify that the provision applies only to the class of licensees of operating reactors that were required by previous versions of § 50.55a to develop and implement a containment inservice inspection program in accordance with Subsection IWE and Subsection IWL, and complete an expedited examination of containment during the 5-year period from September 9, 1996 to September 9, 2001. 10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment Examinations The NRC proposes to revise § 50.55a(b)(2)(viii) by removing the condition for using the 2009 Addenda up to and including the 2013 Edition of Subsection IWL requiring compliance with § 50.55a(b)(2)(viii)(E). 10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth Provision The NRC proposes to add § 50.55a(b)(2)(viii)(H) to require licensees to provide the applicable information specified in paragraphs (b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and (b)(2)(viii)(E)(3) of this section in the ISI Summary Report required by IWA–6000 for each inaccessible concrete surface area evaluated under the new code provision IWL–2512 of the 2009 Addenda up to and including the 2013 Edition. 10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth Provision The NRC proposes to add § 50.55a(b)(2)(viii)(I) containing a new condition requiring the technical evaluation required by IWL–2512(b) of the 2009 Addenda up to and including the 2013 Edition of inaccessible belowgrade concrete surfaces exposed to foundation soil, backfill, or groundwater be performed at periodic intervals not to exceed 5 years. In addition, the licensee must examine representative samples of the exposed portions of the below-grade concrete, when such below-grade concrete is excavated for any reason. The proposed condition would apply only to holders of renewed licenses under 10 CFR part 54 during the period of extended operation (i.e., beyond the PO 00000 Frm 00023 Fmt 4701 Sfmt 4702 56841 expiration date of the original 40-year license) of a renewed license when using IWL–2512(b) of the 2007 Edition with 2009 Addenda through the 2013 Edition. 10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment Examinations The NRC proposes to revise § 50.55a(b)(2)(ix) to continue to apply the existing conditions in § 50.55a(b)(2)(ix)(A)(2), § 50.55a(b)(2)(ix)(B) and § 50.55a(b)(2)(ix)(J) with respect to the metal containment examination requirements in Subsection IWE to the 2009 Addenda up to and including the 2013 Edition and to make minor editorial corrections. 10 CFR 50.55a(b)(2)(ix)(D) Metal Containment Examinations: Fourth Provision The NRC proposes to revise the rule text in § 50.55a(b)(2)(ix)(D) to improve clarity. Paragraphs § 50.55a(b)(2)(ix)(D) and § 50.55a(b)(2)(ix)(D)(1) are combined. The information required to be included in the ISI Summary report is now all on the same paragraph level. No substantive change to the requirements is intended by this revision. 10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance The NRC proposes to revise § 50.55a(b)(2)(x) to clarify that it allows, but does not require, licensees to use the 1994 or the 2008 Edition through the 2009–1a Addenda of NQA–1 when applying the 2009 Addenda and later editions and addenda of the ASME BPV Code, Section XI, up to the 2013 Edition. Licensees are required to meet appendix B of 10 CFR part 50, and NQA–1 is one way of meeting portions of appendix B. A licensee may select any version of NQA–1 that has been approved for use in § 50.55a. NQA–1 provides a method for establishing and implementing a QA program for the design and construction of nuclear power plants and fuel reprocessing plants; however, NQA–1 does not meet all of the requirements of appendix B to 10 CFR part 50. To meet the requirements of appendix B, when using NQA–1 during inservice inspection phase, licensees must address in their quality program description those areas where NQA–1 is insufficient to meet appendix B. Additional guidance and regulatory positions on how to meet appendix B when using NQA–1 is provided in RG 1.28, ‘‘Quality Assurance Criteria (Design and Construction).’’ E:\FR\FM\18SEP5.SGM 18SEP5 56842 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules Section 50.55a(b)(2)(x) clarifies that licensees are required to meet appendix B to 10 CFR part 50 and that the commitments contained in their QA program descriptions that are more stringent than those contained in NQA– 1 or are not addressed in NQA–1 apply to Section XI activities. 10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth Provision The NRC proposes to add § 50.55a(b)(2)(xviii)(D) to provide a new condition prohibiting the use of Appendix VII and subarticle VIII–2200 of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV Code. Licensees would be required to implement Appendix VII and subarticle VIII–2200 of the 2010 Edition of Section XI. 10 CFR 50.55a(b)(2)(xxi)(A) Table IWB– 2500–1 Examination Requirements: First Provision The NRC proposes to revise § 50.55a(b)(2)(xxi)(A) to modify the standard for visual magnification resolution sensitivity and contrast for visual examinations performed on Examination Category B–D components instead of ultrasonic examinations. A visual examination with magnification that has a resolution sensitivity to resolve 0.044 inch (1.1 mm) lower case characters without an ascender or descender (e.g., a, e, n, v), utilizing the allowable flaw length criteria in Table IWB–3512–1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed instead of an ultrasonic examination. This revision removes a requirement that was in addition to ASME BPV Code that required 1-mil wires to be used in licensees’ Sensitivity, Resolution and Contrast Standard targets. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator Preservice Examinations The NRC proposes to add § 50.55a(b)(2)(xxx) to provide a new condition requiring that instead of the preservice inspection requirements of Section XI, IWB–2200(c), a full length examination of 100 percent of the tubing in each newly installed steam generator shall be performed prior to plant startup. These inspections shall be performed with the objective of finding the types of flaws that may potentially be present in the tubes and that may potentially occur during operation. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping Devices The NRC proposes to add § 50.55a(b)(2)(xxxi) to provide a new condition prohibiting the use of mechanical clamping devices in accordance with IWA–4131.1(c) in the 2010 Edition and IWA–4131.1(d) in the 2011 Addenda through 2013 Edition on small item Class 1 piping and portions of a piping system that forms the containment boundary. 10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report Submittal The NRC proposes to add § 50.55a(b)(2)(xxxii) to provide a new condition requiring licensees using the 2010 Edition or later editions and addenda of Section XI to follow the requirements of IWA–6240 of the 2009 addenda of Section XI for the submittal of Preservice and Inservice Summary Reports. 10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed Allowable Pressure The NRC proposes to add § 50.55a(b)(2)(xxxiii) to provide a new condition to prohibit the use of Appendix G Paragraph G–2216 in the 2011 Addenda and later editions and addenda of the ASME BPV Code, Section XI. 10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Disposition of Flaws in Class 3 Components The NRC proposes to add § 50.55a(b)(2)(xxxiv) to provide a new condition to require that when using the 2013 Edition of the ASME BPV Code, Section XI, the licensee shall use the acceptance standards of IWD–3510 for the disposition of flaws in Category D– A components (i.e., welded attachments for vessels, piping, pumps, and valves). 10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 in the KIa and KIc Equations The NRC proposes to add § 50.55a(b)(2)(xxxv) to provide a new condition to specify that when licensees use ASME BPV Code, Section XI, 2013 Edition Appendix A paragraph A–4200, if T0 is available, then RTT0 may be used in place of RTNDT for applications using the KIc equation and the associated KIc curve, but not for applications using the KIa equation and the associated KIa curve. PO 00000 Frm 00024 Fmt 4701 Sfmt 4702 10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of Irradiated Materials The NRC proposes to add § 50.55a(b)(2)(xxxvi) to provide a new condition requiring licensees using ASME BPV Code, Section XI, 2013 Edition, Appendix A, paragraph A– 4400, to obtain NRC approval before using irradiated T0 and the associated RTT0 in establishing fracture toughness of irradiated materials. 10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N–824 The NRC proposes to add § 50.55a(b)(2)(xxxvii) with subparagraphs (A) through (E) to provide a new provision that allows licensees to implement ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ as conditioned by subparagraphs (A) through (E). 10 CFR 50.55a(b)(2)(xxxvii)(A) Section XI Condition: ASME BPV Code Case N– 824 The NRC proposes to add § 50.55a(b)(2)(xxxvii)(A) to add a new condition that requires ultrasonic examinations performed to implement ASME BPV Code Case N–824 to be spatially encoded. 10 CFR 50.55a(b)(2)(xxxvii)(B) Section XI Condition: ASME BPV Code Case N– 824 The NRC proposes to add § 50.55a(b)(2)(xxxvii)(B) to add a new condition that requires that ultrasonic examinations performed to implement ASME BPV Code Case N–824 shall use dual, transmit-receive, refracted longitudinal wave, multi-element phased array search units instead of the requirements of Paragraph 1(c)(1)(–a) of N–824. 10 CFR 50.55a(b)(2)(xxxvii)(C) Section XI Condition: ASME BPV Code Case N– 824 The NRC proposes to add § 50.55a(b)(2)(xxxvii)(C) to add a new condition that requires that ultrasonic examinations performed to implement ASME BPV Code Case N–824 on piping less than or equal to 1.6 inches thick shall use a phased array search unit with a center frequency of 500 kHz to 1 MHz instead of the requirements of Paragraph 1(c)(1)(–c)(–1). 10 CFR 50.55a(b)(2)(xxxvii)(D) Section XI Condition: ASME BPV Code Case N– 824 The NRC proposes to add § 50.55a(b)(2)(xxxvii)(D) to add a new E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules condition that requires that ultrasonic examinations performed to implement ASME BPV Code Case N–824 on piping greater than 1.6 inches thick shall use a phased array search unit with a center frequency of 500 kHz instead of the requirements of Paragraph 1(c)(1)(–c)(– 2). Paragraph 50.55a(b)(3)(i) clarifies that licensees are required to meet appendix B to 10 CFR part 50 and that the commitments contained in their QA program descriptions that are more stringent than those contained in NQA– 1 or are not addressed in NQA–1 apply to OM Code activities. 10 CFR 50.55a(b)(2)(xxxvii)(E) Section XI Condition: ASME BPV Code Case N– 824 10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) Testing The NRC proposes to add § 50.55a(b)(2)(xxxvii)(E) to add a new condition that requires that ultrasonic examinations performed to implement ASME BPV Code Case N–824 shall use a phased array search unit which produces angles from 30 to 70 degrees with a maximum increment of 5 degrees instead of the requirements of Paragraph 1(c)(1)(–d). 10 CFR 50.55a(b)(3) Conditions on ASME OM Code The NRC proposes to revise § 50.55a(b)(3) to require that the 2012 Edition of the ASME OM Code be used during the initial 120-month inservice test interval under § 50.55a(f)(4)(i) and during mandatory 120-month IST program updates under § 50.55a(f)(4)(ii). The proposed revision would also allow users to voluntarily update their IST programs to the 2009 Edition, 2011 Addenda, or 2012 Edition of the ASME OM Code (with the exceptions and conditions specified in this notice) under § 50.55a(f)(4)(iv). mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance The NRC proposes to revise § 50.55a(b)(3)(i) to allow licensees to use the 1983 Edition through the 1994 Edition, 2008 Edition, and 2009–1a Addenda of NQA–1 when using the 1995 Edition through the 2012 Edition of the ASME OM Code. Licensees are required to meet appendix B to 10 CFR part 50, and NQA–1 is one way of meeting portions of appendix B. NQA–1 provides a method for establishing and implementing a QA program for the design and construction of nuclear power plants and fuel reprocessing plants; however, NQA–1 does not meet all of the requirements of appendix B to 10 CFR part 50. To meet the requirements of appendix B, licensees must address in their quality program description those areas where NQA–1 is insufficient to meet appendix B. Regulatory Guide 1.28, ‘‘Quality Assurance Criteria (Design and Construction),’’ provides additional guidance on how to meet appendix B when using NQA–1. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 The NRC proposes to revise § 50.55a(b)(3)(ii) to reflect Appendix III, ‘‘Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,’’ in the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. 10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval 56843 the assumptions in the plant safety analyses. 10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors The NRC proposes to add § 50.55a(b)(3)(iii) to specify that, in addition to complying with the provisions in the OM Code as required with the conditions specified in § 50.55a(b)(3), holders of operating licenses for nuclear power reactors that received construction permits under this part on or after the date 12 months after the effective date of this rulemaking and holders of COLs issued under 10 CFR part 52, whose initial fuel loading occurs on or after the date 12 months after the effective date of this rulemaking, shall also comply with specified conditions, as applicable. The NRC proposes to add § 50.55a(b)(3)(ii)(A) to require that licensees evaluate the adequacy of the diagnostic test interval for each MOV and adjust the interval as necessary, but not later than 5 years or three refueling outages (whichever is longer) from initial implementation of Appendix III of the ASME OM Code. 10 CFR 50.55a(b)(3)(iii)(A) PowerOperated Valves 10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk 10 CFR 50.55a(b)(3)(iii)(B) Check Valves The NRC proposes to add § 50.55a(b)(3)(ii)(B) to require that licensees ensure that the potential increase in core damage frequency and large early release frequency associated with the extension is acceptably small when extending exercise test intervals for high risk MOVs beyond a quarterly frequency. The NRC proposes to add § 50.55a(b)(3)(iii)(B) to require that licensees subject to § 50.55a(b)(3)(iii) perform bi-directional testing of check valves within the IST program where practicable. 10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization The NRC proposes to add § 50.55a(b)(3)(ii)(C) to require, when applying Appendix III to the ASME OM Code, that licensees categorize MOVs according to their safety significance using the methodology described in ASME OM Code Case OMN–3 subject to the conditions discussed in RG 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety evaluation. 10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time The NRC proposes to add § 50.55a(b)(3)(ii)(D) to require, when applying Paragraph III–3600, ‘‘MOV Exercising Requirements,’’ of Appendix III to the OM Code, licensees shall verify that the stroke time of the MOV satisfies PO 00000 Frm 00025 Fmt 4701 Sfmt 4702 The NRC proposes to add § 50.55a(b)(3)(iii)(A) to require that licensees subject to § 50.55a(b)(3)(iii) develop a program to periodically verify the capability of power-operated valves (POVs) to perform their design-basis safety functions. 10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration The NRC proposes to add § 50.55a(b)(3)(iii)(C) to require that licensees subject to § 50.55a(b)(3)(iii) monitor flow-induced vibration (FIV) from hydrodynamic loads and acoustic resonance during preservice testing and inservice testing to identify potential adverse flow effects that might impact components within the scope of the IST program. 10 CFR 50.55a(b)(3)(iii)(D) High Risk Non-Safety Systems The NRC proposes to add § 50.55a(b)(3)(iii)(D) to require that licensees subject to § 50.55a(b)(3)(iii) establish a program to assess the operational readiness of pumps, valves, and dynamic restraints within the scope of the Regulatory Treatment of NonSafety Systems (RTNSS) for applicable reactor designs. The proposed rule language refers to such components using the term, ‘‘high risk non-safety systems.’’ E:\FR\FM\18SEP5.SGM 18SEP5 56844 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules 10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II) The NRC proposes to revise § 50.55a(b)(3)(iv) to specify that Appendix II in the 2003 Addenda through the 2012 Edition of the OM Code is acceptable for use without conditions with the clarifications that (1) the maximum test interval allowed by Appendix II for individual check valves in a group of two valves or more must be supported by periodic testing of a sample of check valves in the group during the allowed interval and (2) the periodic testing plan must be designed to test each valve of a group at approximate equal intervals not to exceed the maximum requirement interval. The conditions currently specified for the use of Appendix II, 1995 Edition with the 1996 and 1997 Addenda, and 1998 Edition through the 2002 Addenda, of the OM Code remain the same in this proposed rule. 10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB The NRC proposes to add § 50.55a(b)(3)(vii) to prohibit the use of Subsection ISTB in the 2011 Addenda to the ASME OM Code. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE The NRC proposes to add § 50.55a(b)(3)(viii) to specify that licensees who wish to implement Subsection ISTE, ‘‘Risk-Informed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,’’ of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition, must first request and obtain NRC approval in accordance with § 50.55a(z) to apply Subsection ISTE on a plant-specific basis as a risk-informed alternative to the applicable IST requirements in the ASME OM Code. 10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF The NRC proposes to add § 50.55a(b)(3)(ix) to specify that licensees applying Subsection ISTF, ‘‘Inservice Testing of Pumps in LightWater Reactor Nuclear Power Plants— Post-2000 Plants,’’ in the 2012 Edition of the OM Code shall satisfy the requirements of Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program,’’ of the OM Code, 2012 Edition. The proposed paragraph will also state that Subsection ISTF, 2011 Addenda, is not acceptable for use. 10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN–20 The NRC proposes to add § 50.55a(b)(3)(x) to allow licensees to VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 implement ASME OM Code Case OMN– 20, ‘‘Inservice Test Frequency,’’ in the ASME OM Code, 2012 Edition. 10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication paragraph (f)(3)(iii)(B) for Class 1 pumps and valves. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME Code and in SRP Section 3.9.6. The NRC proposes to add § 50.55a(b)(3)(xi) to require that licensees supplement the ASME OM Code provisions in Subsection ISTC– 3700, ‘‘Position Verification Testing,’’ as necessary to verify that valve operation is accurately indicated. The ASME OM Code, Subsection ISTC–3700 requires valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. 10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for Operating Plants 10 CFR 50.55a(f): Inservice Testing Requirements 10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements The NRC proposes to revise § 50.55a(f) to clarify that the ASME OM Code includes provisions for preservice testing of components as part of its overall provisions for IST programs. The NRC proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and (g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i), (g)(3)(ii), and (g)(3)(v) to distinguish the requirements for accessibility and preservice examination from those for inservice inspection in § 50.55a(g). No substantive change to the requirements is intended by these revisions. 10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision The NRC proposes to revise § 50.55a(f)(3)(iii)(A) to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision The NRC proposes to revise § 50.55a(f)(3)(iii)(B) to ensure that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First Provision The NRC proposes to revise § 50.55a(f)(3)(iv)(A) to ensure that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(A) for Class 1 pumps and valves. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME Code and in SRP Section 3.9.6. 10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second Provision The NRC proposes to revise § 50.55a(f)(3)(iv)(B) to ensure that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by PO 00000 Frm 00026 Fmt 4701 Sfmt 4702 The NRC proposes to revise § 50.55a(f)(4) to ensure that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME Code and in SRP Section 3.9.6. 10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(1) to require licensees to implement an augmented inservice inspection program for the examination of the RPV upper head penetrations meeting ASME BPV Code Case N–729–4 instead of the previously approved requirements to use ASME BPV Code Case N–729–1, as conditioned by the NRC. 10 CFR 50.55a(g)(6)(ii)(D)(2) Through (5) of the Current Regulation The NRC proposes to remove the conditions in existing § 50.55a(g)(6)(ii)(D)(2) through (5) of the current regulation, inasmuch as these conditions have been included in or reflected in other Code requirements. In their place, the NRC proposes to adopt new conditions in § 50.55a(g)(6)(ii)(D)(2) through (4). 10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I Use The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(2) to require NRC approval prior to implementing Appendix I of ASME BPV Code Case N– 729–4. This requirement is currently located in § 50.55a(g)(6)(ii)(D)(6) for implementation of N–729–1. E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules 10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(3) to add a new condition which requires cold head plants (EDY<8) without PWSCC flaws to perform a bare metal visual examination (VE) each outage a volumetric exam is not performed and allows these plants to extend the bare metal visual inspection frequency from once each refueling outage, as stated in Table 1 of N–729–4, to once every 5 years only if the licensee performed a wetted surface examination of all of the partial penetration welds during the previous volumetric examination. In addition, this new condition clarifies that a bare metal visual examination is not required during refueling outages when a volumetric or surface examination is performed of the partial penetration welds. The condition that is in the current § 50.55a(g)(6)(ii)(D)(3) was incorporated into N–729–4 by the ASME Code committees. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria The NRC proposes to revise § 50.55a(g)(6)(ii)(D)(4) to add a new condition which clarifies that rounded indications found by surface examinations of the partial-penetration or associated fillet welds in accordance with N–729–4 must meet the acceptance criteria for surface examinations of paragraph NB–5352 of ASME Section III of the current edition and addenda for the licensee’s ongoing 10-year inservice inspection interval. The condition that is in the current § 50.55a(g)(6)(ii)(D)(4) was incorporated into N–729–4 by the ASME Code committees. 10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(1) to require licensees to implement an augmented inservice inspection program for the examination of ASME Class 1 piping and nozzle butt welds meeting ASME BPV Code Case N–770–2 instead of the previously approved ASME BPV Code Case N–770–1. Furthermore, the NRC proposes to revise § 50.55a(g)(6)(ii)(F)(1) to update the date of applicability for pressurized water reactors, to note the change to implement ASME BPV Code Case N– 770–2 instead of N–770–1, and to reflect the number of conditions which must be applied. 10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(2) to clarify the VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 requirements for licensees to establish the initial categorization of each weld and modify the wording to reflect the ASME BPV Code Case N–770–2 change in the inspection item category for full structural weld overlays. Additionally, the NRC proposes to add a sentence which clarifies the NRC position that paragraph -1100(e) of ASME BPV Code Case N–770–2 shall not be used to exempt welds that rely on Alloy 82/182 for structural integrity from any requirement of § 50.55a(g)(6)(ii)(F). 10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(3) to clarify the current requirement in this paragraph to complete baseline examinations. Additionally, this condition clarifies that the examination coverage requirements, for a licensee to count previous inspections as baseline examinations, are the same examination coverage requirements described in paragraphs -2500(a) or -2500(b) of ASME BPV Code Case N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(4) to clarify that licensees are required to ensure greater than 90 percent volumetric examination coverage is obtained for circumferential flaws, to continue the restriction on the licensee’s use of paragraph –2500(c) and to continue the restriction that the use of new paragraph –2500(d) of ASME BPV Code Case N–770–2 is not allowed without prior NRC review and approval in accordance with § 50.55a(z), as it would permit a reduction in volumetric examination coverage for circumferential flaws. 10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(5) to add explanatory heading and to make minor editorial corrections. 10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(6) to add explanatory heading. 10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ‘‘t’’ The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(7) to add explanatory heading. PO 00000 Frm 00027 Fmt 4701 Sfmt 4702 56845 10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(8) to continue the current condition located in § 50.55a(g)(6)(ii)(F)(9) which requires that the initial examination of optimized weld overlays (i.e., Inspection Item C–2 of ASME BPV Code Case N–770–2) be performed between the third refueling outage and no later than 10 years after application of the overlay and delete the other current examination requirements for optimized weld overlay examination frequency, as these requirements were included in the revision from N–770–1 to N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(9) to modify the current condition to continue denial of the deferral of the initial inservice examination of uncracked welds mitigated by optimized weld overlays. These welds shall continue to have their initial inservice examinations as prescribed in N–770–1 within 10 years of the application of the optimized weld overlay and not allow deferral of this initial examination. Subsequent inservice examinations may be deferred as allowed by N–770–2. Additionally, the modified condition will delete the current condition on examination requirements for the deferral of welds mitigated by inlay, onlay, stress improvement and optimized weld overlay, as these requirements were, with one exception (i.e., optimized weld overlay), included in the revision from N–770–1 to N–770–2. 10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique The NRC proposes to revise § 50.55a(g)(6)(ii)(F)(10) to modify the current condition to allow the previously prohibited alternate examination requirements of Note (b) of Figure 5(a) of ASME BPV Code Case N–770–1 and N–770–2 and the same requirements in Note 14(b) of Table 1 of ASME BPV Code Case N–770–2 for optimized weld overlays only if the full examination requirements of Note 14(a) of Table 1 of ASME BPV Code Case N– 770–2 cannot be met. 10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel The NRC proposes to add § 50.55a(g)(6)(ii)(F)(11) to provide a new condition requiring licensees to establish a Section XI Appendix VIII qualification requirement for ultrasonic inspection of and through cast stainless steel to meet the examination requirements of paragraph -2500(a) of E:\FR\FM\18SEP5.SGM 18SEP5 56846 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules ASME BPV Code Case N–770–2 by January 1, 2020. 10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage The NRC proposes to add § 50.55a(g)(6)(ii)(F)(12) to provide a new condition that would allow licenses to implement a stress improvement mitigation technique for items containing cast stainless steel that would meet the requirements of Appendix I of ASME BPV Code Case N– 770–2, if the required examination volume can be examined by Appendix VIII procedures to the maximum extent practical including 100 percent of the susceptible material volume. 10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination The NRC proposes to add § 50.55a(g)(6)(ii)(F)(13) to provide a new condition requiring licensees to perform encoded examinations of essentially 100 percent of the inspection surface area when required to perform volumetric examinations of all non-mitigated and cracked mitigated butt welds in accordance with N–770–2. V. Generic Aging Lessons Learned Report mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Background In December 2010, the NRC issued ‘‘Generic Aging Lessons Learned (GALL) Report,’’ NUREG–1801, Revision 2, for applicants to use in preparing their license renewal applications. The GALL Report provides aging management programs (AMPs) that the NRC staff has concluded are sufficient for aging management in accordance with the license renewal rule, as required in 10 CFR 54.21(a)(3). In addition, ‘‘Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,’’ NUREG–1800, Revision 2 was issued in December 2010 to ensure the quality and uniformity of NRC staff reviews of license renewal applications and to present a welldefined basis on which the NRC staff evaluates the applicant’s aging management programs and activities. In April 2011, the NRC also issued ‘‘Disposition of Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG–1801 and NUREG–1800,’’ NUREG–1950, which describes the technical bases for the changes in Revision 2 of the GALL Report and Revision 2 of the SRP for review of license renewal applications. Revision 2 of the GALL Report, in Sections XI.M1, XI.S1, XI.S2, and XI.S3, describes the evaluation and technical VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 bases for determining the sufficiency of ASME BPV Code Subsections IWB, IWC, IWD, IWE, IWF, and IWL for managing aging during the period of extended operation. In addition, many other aging management programs in the GALL Report rely, in part but to a lesser degree, on the requirements specified in the ASME BPV Code, Section XI. Revision 2 of the GALL Report also states that the 1995 Edition through the 2004 Edition of the ASME BPV Code, Section XI, Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as modified and limited by § 50.55a, were found to be acceptable editions and addenda for complying with the requirements of 10 CFR 54.21(a)(3), unless specifically noted in certain sections of the GALL Report. The GALL Report further states that the future Federal Register notices that amend § 50.55a will discuss the acceptability of editions and addenda more recent than the 2004 edition for their applicability to license renewal. In a final rule issued on June 21, 2011 (76 FR 36232), subsequent to Revision 2 of the GALL Report, the NRC also found that the 2004 Edition with the 2005 Addenda through the 2007 Edition with the 2008 Addenda of Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as subject to the conditions in § 50.55a, are acceptable for the AMPs in the GALL Report and the conclusions of the GALL Report remain valid with the augmentations specifically noted in the GALL Report. Evaluation With Respect to Aging Management As part of this rulemaking, the NRC evaluated whether those AMPs in Revision 2 of the GALL Report which rely upon Subsections IWB, IWC, IWD, IWE, IWF, and IWL of Section XI in the editions and addenda of the ASME BPV Code incorporated by reference into § 50.55a, continue to be acceptable if the AMP relies upon the versions of these Subsections in the 2007 Edition with the 2009 Addenda through the 2013 Edition. The NRC finds that the 2007 Edition with the 2009 Addenda through the 2013 Edition of Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as subject to the conditions of this rule, are acceptable for the AMPs in the GALL Report and the conclusions of the GALL Report remain valid with the augmentations specifically noted in the GALL Report. Accordingly, an applicant for license renewal may use, in its plantspecific license renewal application, Subsections IWB, IWC, IWD, IWE, IWF, and IWL of Section XI of the 2007 Edition with the 2009 Addenda through PO 00000 Frm 00028 Fmt 4701 Sfmt 4702 the 2013 Edition of the ASME BPV Code, as subject to the conditions in this rule, without additional justification. Similarly, a licensee approved for license renewal that relied on the GALL AMPs may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of Section XI of the 2007 Edition with the 2009 Addenda through the 2013 Edition of the ASME BPV Code. However, a licensee must assess and follow applicable NRC requirements with regard to changes to its licensing basis. Some of the AMPs in the GALL Report recommend augmentation of certain Code requirements in order to ensure adequate aging management for license renewal. The technical and regulatory aspects of the AMPs for which augmentations are recommended also apply if the editions or addenda from the 2007 Edition with the 2009 Addenda through the 2013 Edition of Section XI of the ASME BPV Code are used to meet the requirements of 10 CFR 54.21(a)(3). The NRC staff evaluated the changes in the 2007 Edition with the 2009 Addenda through the 2013 Edition of Section XI of the ASME BPV Code to determine if the augmentations described in the GALL Report remain necessary; the NRC staff’s evaluation has concluded that the augmentations described in the GALL Report are necessary to ensure adequate aging management. For example, Table IWB– 2500–1, in the 2007 Edition with the 2009 Addenda of ASME BPV Code, Section XI, Subsection IWB, requires surface examination of ASME Code Class 1 branch pipe connection welds less than nominal pipe size (NPS) 4 under Examination Category B–J. However, the NRC staff finds that volumetric or opportunistic destructive examination rather than surface examination is necessary to adequately detect and manage the aging effect due to stress corrosion cracking or thermal, mechanical and vibratory loadings in the components for the period of extended operation. Therefore, GALL Report Section XI.M35, ‘‘One-Time Inspection of ASME Code Class 1 SmallBore Piping,’’ includes the augmentation of the requirements in ASME BPV Code, Section XI, Subsection IWB to perform a one-time inspection of a sample of ASME Code Class 1 piping less than NPS 4 and greater than or equal to NPS 1 using volumetric or opportunistic destructive examination. The GALL Report addresses this augmentation to confirm that there is no need to manage agerelated degradation through periodic volumetric inspections or that an existing AMP (for example, Water E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules Chemistry AMP) is effective to manage the aging effect due to stress corrosion cracking or thermal, mechanical and vibratory loadings for the period of extended operation. A license renewal applicant may either augment its AMPs as described in the GALL Report, or propose alternatives for the NRC to review as part of the applicant’s plantspecific justification for its AMPs. mstockstill on DSK4VPTVN1PROD with PROPOSALS5 VI. Specific Request for Comments The NRC requests specific comments on the following questions: NRC Question 1. NQA–1. The NRC is considering removing the references to versions of NQA–1 older than the 1994 Edition in § 50.55a(b)(1)(iv), § 50.55a(b)(2)(x), and § 50.55a(b)(3)(i). The NRC requests public comment on whether any applicant or licensee is committed to, and is using, a version of NQA–1 older than the 1994 Edition, and if so, what version the applicant or licensee is using. NRC Question 2. ASME BPV Code Case N–824. The NRC is proposing to make ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ acceptable for use with conditions. The use of N–824, as conditioned, is considered a stop-gap improvement until ASME Section XI Appendix VIII Supplement 9 is developed and implemented. The NRC is considering whether ASME BPV Code Case N–824, as conditioned, should be mandatory because of the potential that licensees may continue to use less effective ASME Code Section XI Appendix III techniques for examinations of welds next to CASS material. Should ASME BPV Code Case N–824, as conditioned, be mandatory? What are the possible advantages and disadvantages of making N–824, as conditioned, mandatory? VII. Plain Writing The Plain Writing Act of 2010 (Pub. L. 111–274) requires Federal agencies to write documents in a clear, concise, and well-organized manner. The NRC has written this document to be consistent with the Plain Writing Act as well as the Presidential Memorandum, ‘‘Plain Language in Government Writing,’’ published June 10, 1998 (63 FR 31883). The NRC requests comment on this document with respect to the clarity and effectiveness of the language used. VIII. Voluntary Consensus Standards The National Technology Transfer and Advancement Act of 1995, Public Law 104–113 (NTTAA), and implementing guidance in U.S. Office of Management and Budget (OMB) VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 Circular A–119 (February 10, 1998), requires that Federal agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless using such a standard is inconsistent with applicable law or is otherwise impractical. The NTTAA requires Federal agencies to use industry consensus standards to the extent practical; it does not require Federal agencies to endorse a standard in its entirety. Neither the NTTAA nor Circular A–119 prohibit an agency from adopting a voluntary consensus standard while taking exception to specific portions of the standard, if those provisions are deemed to be ‘‘inconsistent with applicable law or otherwise impractical.’’ Furthermore, taking specific exceptions furthers the Congressional intent of Federal reliance on voluntary consensus standards because it allows the adoption of substantial portions of consensus standards without the need to reject the standards in their entirety because of limited provisions that are not acceptable to the agency. In this rulemaking, the NRC is continuing its existing practice of establishing requirements for the design, construction, operation, inservice inspection (examination) and inservice testing of nuclear power plants by approving the use of the latest editions and addenda of the ASME BPV and OM Codes (ASME Codes) in § 50.55a. The ASME Codes are voluntary consensus standards, developed by participants with broad and varied interests, in which all interested parties (including the NRC and licensees of nuclear power plants) participate. Therefore, the NRC’s incorporation by reference of the ASME Codes is consistent with the overall objectives of the NTTAA and OMB Circular A–119. As discussed in Section III of this statement of considerations, in this proposed rule the NRC is conditioning the use of certain provisions of the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition to the ASME BPV Code, Section III, Division 1 and the ASME BPV Code, Section XI, Division 1, including NQA–1 (with conditions on its use), as well as the 2009 Edition and 2011 Addenda and 2012 Edition to the ASME OM Code and Code Cases N– 770–2, N–729–4, and N–824. In addition, the proposed rule does not adopt (‘‘excludes’’) certain provisions of the ASME Codes and this statement of considerations, and in the regulatory and backfit analysis for this rulemaking. The NRC believes that this proposed rule complies with the NTTAA and OMB Circular A–119 despite these conditions and ‘‘exclusions.’’ PO 00000 Frm 00029 Fmt 4701 Sfmt 4702 56847 If the NRC did not conditionally accept ASME editions, addenda, and code cases, the NRC would disapprove these entirely. The effect would be that licensees and applicants would submit a larger number of requests for use of alternatives under § 50.55a(z), requests for relief under § 50.55a(f) and (g), or requests for exemptions under § 50.12 and/or § 52.7. These requests would likely include broad-scope requests for approval to issue the full scope of the ASME Code editions and addenda which would otherwise be approved as proposed in this rulemaking (i.e., the request would not be simply for approval of a specific ASME Code provision with conditions). These requests would be an unnecessary additional burden for both the licensee and the NRC, inasmuch as the NRC has already determined that the ASME Codes and Code Cases that are the subject of this rulemaking are acceptable for use (in some cases with conditions). For these reasons, the NRC concludes that this proposed rule’s treatment of ASME Code editions and addenda, and code cases and any conditions placed on them does not conflict with any policy on agency use of consensus standards specified in OMB Circular A– 119. The NRC did not identify any other voluntary consensus standards developed by U.S. voluntary consensus standards bodies for use within the U.S. that the NRC could incorporate by reference instead of the ASME Codes. The NRC also did not identify any voluntary consensus standards developed by multinational voluntary consensus standards bodies for use on a multinational basis that the NRC could incorporate by reference instead of the ASME Codes. The NRC identified codes addressing the same subject as the ASME Codes for use in individual countries. At least one country, Korea, directly translated the ASME Code for use in that country. In other countries (e.g., Japan), ASME Codes were the basis for development of the country’s codes, but the ASME Codes were substantially modified to accommodate that country’s regulatory system and reactor designs. Finally, there are countries (e.g., the Russian Federation) where that country’s code was developed without regard to the ASME Code. However, some of these codes may not meet the definition of a voluntary consensus standard because they were developed by the state rather than a voluntary consensus standards body. Evaluation by the NRC of the countries’ codes to determine whether each code provides a comparable or enhanced level of safety E:\FR\FM\18SEP5.SGM 18SEP5 56848 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 when compared against the level of safety provided under the ASME Codes would require a significant expenditure of agency resources. This expenditure does not seem justified, given that substituting another country’s code for the U.S. voluntary consensus standard does not appear to substantially further the apparent underlying objectives of the NTTAA. In summary, this proposed rulemaking satisfies the requirements of the NTTAA and OMB Circular A–119. IX. Incorporation by Reference— Reasonable Availability to Interested Parties The NRC proposes to incorporate by reference seven recent editions and addenda to the ASME codes for nuclear power plants and a standard for quality assurance. The NRC is also proposing to incorporate by reference four ASME code cases. As described in the ‘‘Background’’ and ‘‘Discussion’’ sections of this notice, these materials provide rules for safety governing the design, fabrication, and inspection of nuclear power plant components. The NRC is required by law to obtain approval for incorporation by reference from the Office of the Federal Register (OFR). The OFR’s requirements for incorporation by reference are set forth in 1 CFR part 51. On November 7, 2014, the OFR adopted changes to its regulations governing incorporation by reference (79 FR 66267). The OFR regulations require an agency to include in a proposed rule a discussion of the ways that the materials the agency proposes to incorporate by reference are reasonably available to interested parties or how it worked to make those materials reasonably available to interested parties. The discussion in this section complies with the requirement for proposed rules as set forth in 10 CFR 51.5(a)(1). The NRC considers ‘‘interested parties’’ to include all potential NRC stakeholders, not only the individuals and entities regulated or otherwise subject to the NRC’s regulatory oversight. These NRC stakeholders are not a homogenous group but vary with respect to the considerations for determining reasonable availability. Therefore, the NRC distinguishes between different classes of interested parties for purposes of determining whether the material is ‘‘reasonably available.’’ The NRC considers the following to be classes of interested parties in NRC rulemakings with regard to the material to be incorporated by reference: • Individuals and small entities regulated or otherwise subject to the VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 NRC’s regulatory oversight (this class also includes applicants and potential applicants for licenses and other NRC regulatory approvals) and who are subject to the material to be incorporated by reference by rulemaking. In this context, ‘‘small entities’’ has the same meaning as a ‘‘small entity’’ under 10 CFR 2.810. • Large entities otherwise subject to the NRC’s regulatory oversight (this class also includes applicants and potential applicants for licenses and other NRC regulatory approvals) and who are subject to the material to be incorporated by reference by rulemaking. In this context, ‘‘large entities’’ are those which do not qualify as a ‘‘small entity’’ under 10 CFR 2.810. • Non-governmental organizations with institutional interests in the matters regulated by the NRC. • Other Federal agencies, states, local governmental bodies (within the meaning of 10 CFR 2.315(c)). • Federally-recognized and Staterecognized 3 Indian tribes. • Members of the general public (i.e., individual, unaffiliated members of the public who are not regulated or otherwise subject to the NRC’s regulatory oversight) who may wish to gain access to the materials which the NRC proposes to incorporate by reference by rulemaking in order to participate in the rulemaking. The NRC makes the materials to be incorporated by reference available for inspection to all interested parties, by appointment, at the NRC Technical Library, which is located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301–415–7000; email: Library.Resource@nrc.gov. Interested parties may purchase a copy of the materials from ASME at Three Park Avenue, New York, NY 10016, or at the ASME Web site https:// www.asme.org/shop/standards. The materials are also accessible through third-party subscription services such as IHS (15 Inverness Way East, Englewood, CO 80112; https://global.ihs.com) and Thomson Reuters Techstreet (3916 Ranchero Dr., Ann Arbor, MI 48108; https://www.techstreet.com). The purchase prices for individual documents range from $225 to $720 and the cost to purchase all documents is approximately $9,000. For the class of interested parties constituting members of the general public who wish to gain access to the 3 State-recognized Indian tribes are not within the scope of 10 CFR 2.315(c). However, for purposes of the NRC’s compliance with 1 CFR 51.5, ‘‘interested parties’’ includes a broad set of stakeholders, including State-recognized Indian tribes. PO 00000 Frm 00030 Fmt 4701 Sfmt 4702 materials to be incorporated by reference in order to participate in the rulemaking, the NRC recognizes that the $9,000 cost may be so high that the materials could be regarded as not reasonably available for purposes of commenting on this rulemaking, despite the NRC’s actions to make the materials available at the NRC’s PDR. Accordingly, the NRC sent a letter to the ASME requesting that they consider enhancing public access to these materials during the public comment period (ADAMS Accession No. ML15085A206). In an April 21, 2015, letter to the NRC, the ASME agreed to make the materials available online in a read-only electronic access format during the public comment period (ADAMS Accession No. ML15112A064). Therefore, the seven editions and addenda to the ASME codes for nuclear power plants, the ASME standard for quality assurance, and the four ASME code cases which the NRC proposes to incorporate by reference in this rulemaking are available in read-only format at the ASME Web site https:// go.asme.org/NRC. The NRC concludes that the materials the NRC proposes to incorporate by reference in this rulemaking are reasonably available to all interested parties because the materials are available to all interested parties in multiple ways and in a manner consistent with their interest in the materials. X. Environmental Assessment and Final Finding of No Significant Environmental Impact This proposed rule action is in accordance with the NRC’s policy to incorporate by reference in § 50.55a new editions and addenda of the ASME BPV and OM Codes to provide updated rules for constructing and inspecting components and testing pumps, valves, and dynamic restraints (snubbers) in light-water nuclear power plants. The ASME Codes are national voluntary consensus standards and are required by the NTTAA to be used by government agencies unless the use of such a standard is inconsistent with applicable law or otherwise impractical. The National Environmental Policy Act (NEPA) requires Federal agencies to study the impacts of their ‘‘major Federal actions significantly affecting the quality of the human environment,’’ and prepare detailed statements on the environmental impacts of the proposed action and alternatives to the proposed action (42 U.S.C. Sec. 4332(C); NEPA Sec. 102(C)). The NRC has determined under NEPA, as amended, and the NRC’s E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 regulations in subpart A of 10 CFR part 51, that this proposed rule is not a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement is not required. The rulemaking does not significantly increase the probability or consequences of accidents, no changes are being made in the types of effluents that may be released off-site, and there is no significant increase in public radiation exposure. The NRC estimates the radiological dose to plant personnel performing the inspections required by ASME BPV Code Case N–770–2 would be about 3 rem per plant over a 10-year interval, and a one-time exposure for mitigating welds of about 30 rem per plant. The NRC estimates the radiological dose to plant personnel performing the inspections required by ASME BPV Code Case N–729–4 would be about 3 rem per plant over a 10-year interval and a one-time exposure for mitigating welds of about 30 rem per plant. As required by 10 CFR part 20, and in accordance with current plant procedures and radiation protection programs, plant radiation protection staff will continue monitoring dose rates and would make adjustments in shielding, access requirements, decontamination methods, and procedures as necessary to minimize the dose to workers. The increased occupational dose to individual workers stemming from the ASME BPV Code Case N–770–2 and N–729–4 inspections must be maintained within the limits of 10 CFR part 20 and as low as reasonably achievable. Therefore, the NRC concludes that the increase in occupational exposure would not be significant. The proposed rule does not involve non-radiological plant effluents and has no other environmental impact. Therefore, no significant nonradiological impacts are associated with this action. The determination of this environmental assessment is that there will be no significant off-site impact to the public from this action. XI. Paperwork Reduction Act Statement This proposed rule contains new or amended collections of information subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This proposed rule has been submitted to the Office of Management and Budget for review and approval of the information collections. Type of submission, new or revision: Revision. The title of the information collection: Domestic Licensing of Production and Utilization Facilities: Incorporation by VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 Reference of American Society of Mechanical Engineers Codes and Code Cases. The form number if applicable: Not applicable. How often the collection is required or requested: On occasion. Who will be required or asked to respond: Power reactor licensees and applicants for power reactors under construction. An estimate of the number of annual responses: 320. The estimated number of annual respondents: 104. An estimate of the total number of hours needed annually to comply with the information collection requirement or request: 121,600. Abstract: This proposed rule is the latest in a series of rulemakings to amend the NRC’s regulations to incorporate by reference revised and updated ASME codes for nuclear power plants. The number of operating nuclear power plants has decreased and the NRC has increased its estimate of the burden associated with developing alternative requests. Overall, the reporting burden for 10 CFR 50.55a has increased. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues: 1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility? 2. Is the estimate of the burden of the proposed information collection accurate? 3. Is there a way to enhance the quality, utility, and clarity of the information to be collected? 4. How can the burden of the proposed information collection on respondents be minimized, including the use of automated collection techniques or other forms of information technology? A copy of the OMB clearance package and proposed rule is available in ADAMS (Accession Nos. ML14141A281 and ML14258B191) or may be viewed free of charge at the NRC’s PDR, One White Flint North, 11555 Rockville Pike, Room O–1 F21, Rockville, MD 20852. You may obtain information and comment submissions related to the OMB clearance package by searching on https://www.regulations.gov under Docket ID NRC–2011–0088. You may submit comments on any aspect of these proposed information collection(s), including suggestions for PO 00000 Frm 00031 Fmt 4701 Sfmt 4702 56849 reducing the burden and on the previously stated issues, by the following methods: • Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC–2011–0088. • Mail comments to: FOIA, Privacy, and Information Collections Branch, Office of Information Services, Mail Stop: T–5 F53, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001 or to Vlad Dorjets, Desk Officer, Office of Information and Regulatory Affairs (3150–0011), NEOB–10202, Office of Management and Budget, Washington, DC 20503; telephone 202– 395–7315, email: oira_submission@ omb.eop.gov. Submit comments by October 19, 2015. Comments received after this date will be considered if it is practical to do so, but the NRC staff is able to ensure consideration only for comments received on or before this date. Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number. XII. Regulatory Analysis: Availability The NRC has prepared a draft regulatory analysis on this proposed rule. The analysis examines the costs and benefits of the alternatives considered by the Commission. The NRC requests public comments on the draft regulatory analysis. Comments on the draft analysis may be submitted to the NRC by any method provided in the ADDRESSES section of this notice. XIII. Backfitting and Issue Finality Introduction The NRC’s Backfit Rule in § 50.109 states that the NRC shall require the backfitting of a facility only when it finds the action to be justified under specific standards stated in the rule. Section 50.109(a)(1) defines backfitting as the modification of or addition to systems, structures, components, or design of a facility; the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct, or operate a facility. Any of these modifications or additions may result from a new or amended provision in the NRC’s rules or the imposition of a regulatory position interpreting the NRC’s rules that is either new or different from a previously applicable NRC position after issuance of the construction permit E:\FR\FM\18SEP5.SGM 18SEP5 56850 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 or the operating license or the design approval. Section 50.55a requires nuclear power plant licensees to: • Construct ASME BPV Code Class 1, 2, and 3 components in accordance with the rules provided in Section III, Division 1, of the ASME BPV Code (‘‘Section III’’). • Inspect Class 1, 2, 3, Class MC, and Class CC components in accordance with the rules provided in Section XI, Division 1, of the ASME BPV Code (‘‘Section XI’’). • Test Class 1, 2, and 3 pumps, valves, and dynamic restraints (snubbers) in accordance with the rules provided in the ASME OM Code. This rulemaking proposes to incorporate by reference the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition to the ASME BPV Code, Section III, Division 1 and ASME BPV Code, Section XI, Division 1, including NQA–1 (with conditions on its use), as well as the 2009 Edition and 2011 Addenda and 2012 Edition to the ASME OM Code and Code Cases N– 770–2 and N–729–4. The ASME BPV and OM codes are national consensus standards developed by participants with broad and varied interests, in which all interested parties (including the NRC and utilities) participate. A consensus process involving a wide range of stakeholders is consistent with the National Technology Transfer and Advancement Act, inasmuch as the NRC has determined that there are sound regulatory reasons for establishing regulatory requirements for design, maintenance, ISI, and IST by rulemaking. The process also facilitates early stakeholder consideration of backfitting issues. Thus, the NRC believes that the NRC need not address backfitting with respect to the NRC’s general practice of incorporating by reference updated ASME Codes. Overall Backfitting Considerations: Section III of the ASME BPV Code Incorporation by reference of more recent editions and addenda of Section III of the ASME BPV Code does not affect a plant that has received a construction permit or an operating license or a design that has been approved. This is because the edition and addenda to be used in constructing a plant are, under § 50.55a, determined based on the date of the construction permit, and are not changed thereafter, except voluntarily by the licensee. The incorporation by reference of more recent editions and addenda of Section III ordinarily applies only to applicants after the effective date of the final rule VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 incorporating these new editions and addenda. Thus, incorporation by reference of a more recent edition and addenda of Section III does not constitute ‘‘backfitting’’ as defined in § 50.109(a)(1). Overall Backfitting Considerations: Section XI of the ASME BPV Code and the ASME OM Code Incorporation by reference of more recent editions and addenda of Section XI of the ASME BPV Code and the ASME OM Code affects the ISI and IST programs of operating reactors. However, the Backfit Rule generally does not apply to incorporation by reference of later editions and addenda of the ASME BPV Code (Section XI) and OM Code. As previously mentioned, the NRC’s longstanding regulatory practice has been to incorporate later versions of the ASME Codes into § 50.55a. Under § 50.55a, licensees shall revise their ISI and IST programs every 120 months to the latest edition and addenda of Section XI of the ASME BPV Code and the ASME OM Code incorporated by reference into § 50.55a 12 months before the start of a new 120-month ISI and IST interval. Thus, when the NRC approves and requires the use of a later version of the Code for ISI and IST, it is implementing this longstanding regulatory practice and requirement. Other circumstances where the NRC does not apply the Backfit Rule to the approval and requirement to use later Code editions and addenda are as follows: 1. When the NRC takes exception to a later ASME BPV Code or OM Code provision but merely retains the current existing requirement, prohibits the use of the later Code provision, limits the use of the later Code provision, or supplements the provisions in a later Code. The Backfit Rule does not apply because the NRC is not imposing new requirements. However, the NRC explains any such exceptions to the Code in the Statement of Considerations and regulatory analysis for the rule. 2. When an NRC exception relaxes an existing ASME BPV Code or OM Code provision but does not prohibit a licensee from using the existing Code provision. The Backfit Rule does not apply because the NRC is not imposing new requirements. 3. Modifications and limitations imposed during previous routine updates of § 50.55a have established a precedent for determining which modifications or limitations are backfits, or require a backfit analysis (e.g., final rule dated September 10, 2008 [73 FR 52731], and a correction dated October 2, 2008 [73 FR 57235]). The application PO 00000 Frm 00032 Fmt 4701 Sfmt 4702 of the backfit requirements to modifications and limitations in the current rule are consistent with the application of backfit requirements to modifications and limitations in previous rules. The incorporation by reference and adoption of a requirement mandating the use of a later ASME BPV Code or OM Code may constitute backfitting in some circumstances. In these cases, the NRC would perform a backfit analysis or documented evaluation in accordance with § 50.109. These include the following: 1. When the NRC endorses a later provision of the ASME BPV Code or OM Code that takes a substantially different direction from the existing requirements, the action is treated as a backfit (e.g., 61 FR 41303 [August 8, 1996]). 2. When the NRC requires implementation of a later ASME BPV Code or OM Code provision on an expedited basis, the action is treated as a backfit. This applies when implementation is required sooner than it would be required if the NRC simply endorsed the Code without any expedited language (e.g., 64 FR 51370 [September 22, 1999]). 3. When the NRC takes an exception to an ASME BPV Code or OM Code provision and imposes a requirement that is substantially different from the existing requirement as well as substantially different from the later Code (e.g., 67 FR 60529 [September 26, 2002]). Detailed Backfitting Discussion: Proposed Changes Beyond Those Necessary To Incorporate by Reference the New ASME BPV and OM Code Provisions This section discusses the backfitting considerations for all the proposed changes to § 50.55a that go beyond the minimum changes necessary and required to adopt the new ASME Code Addenda into § 50.55a. ASME BPV Code, Section III 1. Revise § 50.55a(b)(1)(ii), ‘‘Weld leg dimensions,’’ to clarify rule language and add Table 1, which clarifies prohibited Section III provisions in tabular form for welds with leg size less than 1.09 tn. This proposed change would not alter the original intent of this requirement and, therefore, would not impose a new requirement. Therefore, this proposed change is not a backfit. 2. Revise § 50.55a(b)(1)(iv), ‘‘Section III condition: Quality assurance,’’ to require that when applying editions and addenda later than the 1989 Edition of E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Section III, the requirements of NQA–1, 1983 Edition through the 1994 Edition, 2008 Edition, and the 2009–1a Addenda are acceptable for use, provided that the edition and addenda of NQA–1 specified in either NCA–4000 or NCA– 7000 is used in conjunction with the administrative, quality and technical provisions contained in the edition and addenda of Section III being used. This proposed revision clarifies the current requirements, and is considered to be consistent with the meaning and intent of the current requirements, and therefore is not considered to result in a change in requirements. Therefore, this proposed change is not a backfit. 3. Add a new proposed condition as § 50.55a(b)(1)(viii), ‘‘Use of ASME Certification Marks,’’ to allow licensees to use either the ASME BPV Code Symbol Stamp or ASME Certification Mark with the appropriate certification designator and class designator as specified in the 2013 Edition through the latest edition and addenda incorporated by reference in 10 CFR 50.55a. This proposed condition would not result in a change in requirements previously approved in the Code and, therefore, is not a backfit. ASME BPV Code, Section XI 1. Revise § 50.55a(b)(2)(vi), ‘‘Effective Edition and Addenda of Subsection IWE and Subsection IWL, Section XI;’’ to clarify that the provision applies only to the class of licensees of operating reactors that were required by previous versions of § 50.55a to develop, implement a containment inservice inspection program in accordance with Subsection IWE and Subsection IWL, and complete an expedited examination of containment during the 5-year period from September 9, 1996, to September 9, 2001. This proposed revision clarifies the current requirements, is considered to be consistent with the meaning and intent of the current requirements, and is not considered to result in a change in requirements. Therefore, this proposed change is not a backfit. 2. Revise § 50.55a(b)(2)(viii), ‘‘Examination of concrete containments,’’ so that when using the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL, the conditions in 10 CFR 50.55a(b)(2)(viii)(E) do not apply, but the proposed conditions in new 10 CFR 50.55a(b)(2)(viii)(H) and 10 CFR 50.55a(b)(2)(viii)(I) do apply. This proposed revision would not require 10 CFR 50.55a(b)(2)(viii)(E) to be used when following the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL because most of its requirements have been included in VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 IWL–2512, ‘‘Inaccessible Areas.’’ Therefore, this proposed change is not a backfit because the requirements have not changed. The revision to add the condition in 10 CFR 50.55a(b)(2)(viii)(H) captures the reporting requirements of the current 10 CFR 50.55a(b)(2)(viii)(E) which were not included in IWL–2512. Therefore, this proposed change is not a backfit because the requirements have not changed. The revision to add the condition in 10 CFR 50.55a(b)(2)(viii)(I) addresses a new code provision in IWL– 2512(b) for evaluation of below-grade concrete surfaces during the period of extended operation of a renewed license. The condition assures consistency with the GALL Report and applies to plants going forward using the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWL. The requirements would remain unchanged from those of the GALL Report and, therefore, this change is not a backfit. 3. Revise § 50.55a(b)(2)(ix), ‘‘Examination of metal containments,’’ to extend the applicability of the existing conditions in § 50.55a(b)(2)(ix)(A)(2), § 50.55a(b)(2)(ix)(B), and § 50.55a(b)(2)(ix)(J) to the 2007 Edition with 2009 Addenda through the 2013 Edition of Subsection IWE. This proposed condition would not result in a change to current requirements, and is therefore not a backfit. 4. Revise § 50.55a(b)(2)(x), ‘‘Section XI condition: Quality assurance,’’ to require that when applying the editions and addenda later than the 1989 Edition of ASME BPV Code, Section XI, the requirements of NQA–1, 1983 Edition through the 1994 Edition, the 2008 Edition, and the 2009–1a Addenda specified in either IWA–1400 or Table IWA 1600–1, ‘‘Referenced Standards and Specifications,’’ of that edition and addenda of Section XI are acceptable for use, provided the licensee uses its appendix B to 10 CFR part 50 quality assurance program in conjunction with Section XI requirements. This proposed revision clarifies the current requirements, which the NRC considers to be consistent with the meaning and intent of the current requirements. Therefore, the NRC does not consider the clarification to be a change in requirements. Therefore, this proposed change is not a backfit. 5. Add a new proposed condition as § 50.55a(b)(2)(xviii)(D), ‘‘NDE personnel certification: Fourth provision;’’ to prohibit the use of Appendix VII and subarticle VIII–2200 of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV Code. Licensees would be required to implement PO 00000 Frm 00033 Fmt 4701 Sfmt 4702 56851 Appendix VII and subarticle VIII–2200 of the 2010 Edition of Section XI. This condition does not constitute a change in NRC position because the use of the subject provisions is not currently allowed by § 50.55a. Therefore, the addition of this new proposed condition is not a backfit. 6. Revise § 50.55a(b)(2)(xxi)(A), ‘‘Table IWB–2500–1 examination requirements; First provision,’’ to modify the standard for visual magnification resolution sensitivity and contrast for visual examinations of Examination Category B–D components, making the rule conform with ASME BPV Code, Section XI requirements for VT–1 examinations. This proposed revision removes a condition that was in addition to the ASME Code requirements and does not impose a new requirement. Therefore, this change is not a backfit. 7. Add a new proposed condition as § 50.55a(b)(2)(xxx), ‘‘Steam Generator Preservice Examinations;’’ to require that instead of the preservice inspection requirements of Section XI, IWB– 2200(c), a full length examination of 100 percent of the tubing in each newly installed steam generator shall be performed prior to plant startup. This proposed condition provides a clarification consistent with industry guidelines and the NRC staff position in SRP Section 5.4.2.2. Therefore, the addition of this new proposed condition is not a backfit. 8. Add a new proposed condition as § 50.55a(b)(2)(xxxi), ‘‘Mechanical clamping devices;’’ to prohibit the use of mechanical clamping devices in accordance with IWA–4131.1(c) in the 2010 Edition and IWA–4131.1(d) in the 2011 Addenda through 2013 Edition on small item Class 1 piping and portions of a piping system that forms the containment boundary. This condition does not constitute a change in NRC position and would not affect licensees because the use of the subject provisions is not currently allowed by § 50.55a. Therefore, the addition of this new proposed condition is not a backfit. 9. Add a new proposed condition as § 50.55a(b)(2)(xxxii), ‘‘Summary Report submittal;’’ to clarify that licensees using the 2010 Edition or later editions and addenda of Section XI must continue to submit to the NRC the Preservice and Inservice Summary Reports required by IWA–6240 of the 2009 addenda of Section XI. This proposed condition would not result in a change in NRC’s requirements insomuch as these reports have been required in the 2009 Addenda of Section XI and all previous editions and E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56852 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules addenda. Therefore, the addition of this new proposed condition is not a backfit. 10. Add a new proposed condition as § 50.55a(b)(2)(xxxiii), ‘‘Risk-Informed allowable pressure;’’ to prohibit the use of ASME BPV Code, Section XI, Appendix G, Paragraph G–2216. The use of Paragraph G–2216 is not currently allowed by § 50.55a. Therefore, the proposed condition does not constitute a new or changed NRC position on the lack of acceptability of Paragraph G–2216. Therefore, the addition of this new proposed condition is not a backfit. 11. Add a new proposed condition as § 50.55a(b)(2)(xxxiv), ‘‘Disposition of flaws in Class 3 components;’’ to require that when using the 2013 Edition of the ASME BPV Code, Section XI, the licensee shall use the acceptance standards of IWD–3510 for the disposition of flaws in Category D–A components. The condition is imposed to provide clarification and consistency in requirements between IWD–3410 and IWD–3510. This proposed change would not alter the original intent of this requirement and, therefore, would not impose a new requirement. This proposed change is not a backfit. 12. Add a new proposed condition as § 50.55a(b)(2)(xxxv), ‘‘Use of RTT0 in the KIa and KIc equations;’’ to specify that when licensees use ASME BPV Code, Section XI 2013 Edition Nonmandatory Appendix A paragraph A–4200, if T0 is available, then RTT0 may be used in place of RTNDT for applications using the KIc equation and the associated KIc curve, but not for applications using the KIa equation and the associated KIa curve. Conditions on the use of ASME BPV Code, Section XI, Nonmandatory Appendices do not constitute backfitting inasmuch as those provisions apply to voluntary actions initiated by the licensee to use the ‘‘nonmandatory compliance’’ provisions in these Appendices of the proposed rule. 13. Add a new proposed condition as § 50.55a(b)(2)(xxxvi), ‘‘Fracture toughness of irradiated materials;’’ to require licensees using ASME BPV Code, Section XI 2013 Edition Nonmandatory Appendix A paragraph A–4400, to obtain NRC approval before using irradiated T0 and the associated RTT0 in establishing fracture toughness of irradiated materials. Conditions on the use of ASME BPV Code, Section XI, Nonmandatory Appendices do not constitute backfitting inasmuch as those provisions apply to voluntary actions initiated by the licensee to use the ‘‘nonmandatory compliance’’ provisions in these Appendices of the proposed rule. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 14. Add a new proposed condition as § 50.55a(b)(2)(xxxvii), ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ to allow the use of the code case as conditioned. Conditions on the use of ASME BPV Code Case N–824 do not constitute backfitting, inasmuch as the use of this code case is not required by the NRC but instead is an alternative which may be voluntarily used by the licensee (i.e., a ‘‘voluntary alternative’’). ASME OM Code 1. Add a new proposed condition as § 50.55a(b)(3)(ii)(A) to require that licensees evaluate the adequacy of the diagnostic test interval for each MOV and adjust the interval as necessary, but not later than 5 years or three refueling outages (whichever is longer) from initial implementation of Appendix III of the ASME OM Code. This proposed condition represents an exception to a later OM Code provision but merely retains the current NRC requirement in RG 1.192, and is therefore not a backfit because the NRC is not imposing a new requirement. 2. Add a new proposed condition as § 50.55a(b)(3)(ii)(B) to require that licensees ensure that the potential increase in core damage frequency and large early release frequency associated with the extension is acceptably small when extending exercise test intervals for high risk MOVs beyond a quarterly frequency. This proposed condition represents an exception to a later OM Code provision but merely retains the current NRC requirement in RG 1.192, and is therefore not a backfit because the NRC is not imposing a new requirement. 3. Add a new proposed condition as § 50.55a(b)(3)(ii)(C) to require, when applying Appendix III to the ASME OM Code, that licensees categorize MOVs according to their safety significance using the methodology described in ASME OM Code Case OMN–3 subject to the conditions discussed in RG 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety evaluation. This proposed condition represents an exception to a later OM Code provision but merely retains the current NRC requirement in RG 1.192, and is therefore not a backfit because the NRC is not imposing a new requirement. 4. Add a new proposed condition as § 50.55a(b)(3)(ii)(D) to require that, when applying Paragraph III–3600, ‘‘MOV Exercising Requirements,’’ of Appendix III to the OM Code, licensees PO 00000 Frm 00034 Fmt 4701 Sfmt 4702 shall verify that the stroke time of the MOV satisfies the assumptions in the plant safety analyses. This proposed condition retains the MOV stroke time requirement that was specified in previous editions and addenda of the ASME OM Code. The retention of this requirement is not a backfit. 5. Add new proposed conditions as § 50.55a(b)(3)(iii)(A) through § 50.55a(b)(3)(iii)(D), ‘‘OM condition: New Reactors;’’ to apply specific conditions for IST programs applicable to licensees of new nuclear power plants in addition to the provisions of the ASME OM Code as incorporated by reference with conditions in § 50.55a. Licensees of ‘‘new reactors’’ are, as identified in the proposed paragraph: (i) Holders of operating licenses for nuclear power reactors that received construction permits under this part on or after the date 12 months after the effective date of this rulemaking and (ii) holders of COLs issued under 10 CFR part 52, whose initial fuel loading occurs on or after the date 12 months after the effective date of this rulemaking. This implementation schedule for new reactors is consistent with the NRC regulations in § 50.55a(f)(4)(i). These proposed conditions represent an exception to a later OM Code provision but merely retain the current NRC requirement, and are therefore not a backfit because the NRC is not imposing a new requirement. 6. Revise § 50.55a(b)(3)(iv), ‘‘OM condition: Check valves (Appendix II),’’ to specify that Appendix II, ‘‘Check Valve Condition Monitoring Program,’’ of the OM Code, 2003 Addenda through the 2012 Edition, is acceptable for use without conditions with the clarifications that (1) the maximum test interval allowed by Appendix II for individual check valves in a group of two valves or more must be supported by periodic testing of a sample of check valves in the group during the allowed interval and (2) the periodic testing plan must be designed to test each valve of a group at approximate equal intervals not to exceed the maximum requirement interval. The regulation is being revised to extend the applicability of this existing NRC condition on the OM Code to the 2012 Edition of the OM Code. This does not represent a change in the NRC’s position that the condition is needed with respect to the OM Code. Therefore, this proposed condition is not a backfit. 7. Add a new proposed condition as § 50.55a(b)(3)(vii), ‘‘OM condition: Subsection ISTB;’’ to prohibit the use of Subsection ISTB in the 2011 Addenda to the ASME OM Code because the complete set of planned Code E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules modifications to support the changes to the comprehensive pump test acceptance criteria was not made in that addenda. This proposed condition represents an exception to a later OM Code provision but merely limits the use of the later Code provision, and is therefore not a backfit because the NRC is not imposing a new requirement. 8. Add a new proposed condition as § 50.55a(b)(3)(viii), ‘‘OM condition: Subsection ISTE;’’ to allow licensees to implement Subsection ISTE, ‘‘RiskInformed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,’’ in the ASME OM Code, 2009 Edition, 2011 Addenda and 2012 Edition, where the licensee has obtained authorization to implement Subsection ISTE as an alternative to the applicable IST requirements in the ASME OM Code on a case-by-case basis in accordance with § 50.55a(z). This proposed condition represents an exception to a later OM Code provision but merely limits the use of the later Code provision, and is therefore not a backfit because the NRC is not imposing a new requirement. 9. Add a new proposed condition as § 50.55a(b)(3)(ix), ‘‘OM Condition: Subsection ISTF;’’ to specify that licensees applying Subsection ISTF, 2012 Edition, shall satisfy the requirements of Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program,’’ of the ASME OM Code, 2012 Edition. The proposed condition also specifies that Subsection ISTF, 2011 Addenda, is not acceptable for use. This proposed condition represents an exception to a later OM Code provision but merely limits the use of the later Code provision, and is therefore not a backfit because the NRC is not imposing a new requirement. 10. Add a new proposed condition as § 50.55a(b)(3)(x), ‘‘OM condition: ASME OM Code Case OMN–20,’’ to allow licensees to implement ASME OM Code Case OMN–20, ‘‘Inservice Test Frequency,’’ in the ASME OM Code, 2012 Edition. This proposed condition allows voluntary action initiated by the licensee to use the code case and is, therefore, not a backfit. 11. Add a new proposed condition as § 50.55a(b)(3)(xi), ‘‘OM condition: Valve Position Indication,’’ to specify that when implementing ASME OM Code, Subsection ISTC–3700, ‘‘Position Verification Testing,’’ licensees shall supplement the ASME OM Code provisions as necessary to verify that valve operation is accurately indicated. This proposed condition clarifies the current requirements, and is considered to be consistent with the meaning and intent of the current requirements, and VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 therefore is not considered to result in a change in requirements. As such, this proposed condition is not a backfit. 12. Revise § 50.55a(f), ‘‘Inservice testing requirements,’’ to clarify that the ASME OM Code includes provisions for preservice testing of components as part of its overall provisions for IST programs. No expansion of IST program scope is intended by this clarification. This proposed condition would not result in a change in requirements previously approved in the Code and is, therefore, not a backfit. 13. Revise § 50.55a(f)(3)(iii)(A), ‘‘Class 1 pumps and valves: First provision,’’ to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. This proposed condition would not result in a change in requirements previously approved in the Code and is, therefore, not a backfit. 14. Revise § 50.55a(f)(3)(iii)(B), ‘‘Class 1 pumps and valves: Second provision,’’ to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. This proposed condition would not result in a change in requirements previously approved in the Code and is, therefore, not a backfit. 15. Revise § 50.55a(f)(3)(iv)(A), ‘‘Class 2 and 3 pumps and valves: First provision;’’ to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(A) for Class 1 pumps and valves. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. This proposed condition would not result in a change in requirements previously approved in the Code and is, therefore, not a backfit. 16. Revise § 50.55a(f)(3)(iv)(B), ‘‘Class 2 and 3 pumps and valves: Second provision,’’ to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(B) for Class 1 pumps and valves. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. This proposed condition would not result in a change in requirements previously approved in the Code, and is therefore not a backfit. PO 00000 Frm 00035 Fmt 4701 Sfmt 4702 56853 17. Revise § 50.55a(f)(4), ‘‘Inservice testing standards for operating plants;’’ to state that the paragraph is applicable to pumps and valves that are within the scope of the ASME OM Code. This will align the scope of pumps and valves for inservice testing with the scope defined in the ASME OM Code and in SRP Section 3.9.6. This proposed condition would not result in a change in requirements previously approved in the Code, and is therefore not a backfit. ASME BPV Code Case N–729–4 Revise § 50.55a(g)(6)(ii)(D), ‘‘Reactor vessel head inspections’’: On June 22, 2012, the ASME approved the fourth revision of ASME BPV Code Case N–729, (N–729–4). The NRC proposes to update the requirements of § 50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV Code Case N–729–4, with conditions. The ASME BPV Code Case N–729–4 contains similar requirements as N–729–1; however, N–729–4 also contains new requirements to address previous NRC conditions, including changes to inspection frequency and qualifications. The new NRC conditions on the use of ASME BPV Code Case N– 729–4 address operational experience, clarification of implementation, and the use of alternatives to the code case. The current regulatory requirements for the examination of pressurized water reactor upper RPV heads that use nickel-alloy materials are provided in § 50.55a(g)(6)(ii)(D). This section was first created by rulemaking, dated September 10, 2008, (73 FR 52730) to require licensees to implement ASME BPV Code Case N–729–1, with conditions, instead of the inspections previously required by the ASME BPV Code, Section XI. The action did constitute a backfit; however, NRC concluded that imposition of ASME BPV Code Case N–729–1, as conditioned, constituted an adequate protection backfit. The GDC for nuclear power plants (appendix A to 10 CFR part 50) or, as appropriate, similar requirements in the licensing basis for a reactor facility, provide bases and requirements for NRC assessment of the potential for, and consequences of, degradation of the reactor coolant pressure boundary (RCPB). The applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC 31 (Fracture Prevention of Reactor Coolant Pressure Boundary), and GDC 32 (Inspection of Reactor Coolant Pressure Boundary). General Design Criterion 14 specifies that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56854 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules rapidly propagating failure, and of gross rupture. General Design Criterion 31 specifies that the probability of rapidly propagating fracture of the RCPB be minimized. General Design Criterion 32 specifies that components that are part of the RCPB have the capability of being periodically inspected to assess their structural and leak tight integrity. The NRC concludes that ASME BPV Code Case N–729–4, as conditioned, shall be mandatory in order to ensure that the requirements of the GDC are satisfied. Imposition of ASME BPV Code Case N–729–4, with conditions, ensures that the ASME Code-allowable limits will not be exceeded, leakage will likely not occur and potential flaws will be detected before they challenge the structural or leak tight integrity of the reactor pressure vessel upper head within current nondestructive examination limitations. The NRC concludes that the regulatory framework for providing adequate protection of public health and safety is accomplished by the incorporation of ASME BPV Code Case N–729–4 into § 50.55a, as conditioned. All current licensees of U.S. pressurized water reactors will be required to implement ASME BPV Code Case N–729–4, as conditioned. The Code Case provisions on examination requirements for reactor pressure vessel upper heads are essentially the same as those established under ASME BPV Code Case N–729–1, as conditioned. One exception is the condition in § 50.55a(g)(6)(ii)(D)(3), which will require, for upper heads with Alloy 600 penetration nozzles, that bare metal visual examinations be performed each outage in accordance with Table 1 of ASME BPV Code Case N–729–4. Accordingly, the NRC imposition of the ASME BPV Code Case N–729–4, as conditioned, may be deemed to be a modification of the procedures to operate a facility resulting from the imposition of the new regulation, and as such, this rulemaking provision may be considered backfitting under § 50.109(a)(1). The NRC continues to find that inspections of reactor pressure vessel upper heads, their penetration nozzles, and associated partial penetration welds are necessary for adequate protection of public health and safety and that the requirements of ASME BPV Code Case N–729–4, as conditioned, represent an acceptable approach, developed, in part, by a voluntary consensus standards organization for performing future inspections. The NRC concludes that approval of ASME BPV Code Case N– 729–4, as conditioned, by incorporation by reference of the Code Case into § 50.55a, is necessary to ensure that the VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 facility provides adequate protection to the health and safety of the public and constitutes a redefinition of the requirements necessary to provide reasonable assurance of adequate protection of public health and safety. Therefore, a backfit analysis need not be prepared for this portion of the proposed rule in accordance with § 50.109(a)(4)(ii) and § 50.109(a)(4)(iii). ASME BPV Code Case N–770–2 Revise § 50.55a(g)(6)(ii)(F), ‘‘Examination requirements for Class 1 piping and nozzle dissimilar metal butt welds’’: On June 9, 2011, the ASME approved the second revision of ASME BPV Code Case N–770, (N–770–2). The NRC proposes to update the requirements of § 50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code Case N– 770–2, with conditions. The ASME BPV Code Case N–770–2 contains similar baseline and ISI requirements for unmitigated nickel-alloy butt welds, and preservice and ISI requirements for mitigated butt welds as N–770–1. However, N–770–2 also contains new requirements for optimized weld overlays, a specific mitigation technique and volumetric inspection coverage. Further, the NRC conditions on the use of ASME BPV Code Case N–770–2 have been modified to address the changes in the code case, clarify inspection coverage requirements and require the development of inspection qualifications to allow complete weld inspection coverage in the future. The current regulatory requirements for the examination of ASME Class 1 piping and nozzle dissimilar metal butt welds that use nickel-alloy materials is provided in § 50.55a(g)(6)(ii)(F). This section was first created by rulemaking, dated June 21, 2011 (76 FR 36232), to require licensees to implement ASME BPV Code Case N–770–1, with conditions. The NRC added § 50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code Case N– 770–1, with conditions, instead of the inspections previously required by the ASME BPV Code, Section XI. The action did constitute a backfit; however, the NRC concluded that imposition of ASME BPV Code Case N–770–1, as conditioned, constituted an adequate protection backfit. The GDC for nuclear power plants (appendix A to 10 CFR part 50) or, as appropriate, similar requirements in the licensing basis for a reactor facility, provide bases and requirements for NRC assessment of the potential for, and consequences of, degradation of the RCPB. The applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), PO 00000 Frm 00036 Fmt 4701 Sfmt 4702 GDC 31 (Fracture Prevention of Reactor Coolant Pressure Boundary) and GDC 32 (Inspection of Reactor Coolant Pressure Boundary). General Design Criterion 14 specifies that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. General Design Criterion 31 specifies that the probability of rapidly propagating fracture of the RCPB be minimized. General Design Criterion 32 specifies that components that are part of the RCPB have the capability of being periodically inspected to assess their structural and leak tight integrity. The NRC concludes that ASME BPV Code Case N–770–2, as conditioned, must be imposed in order to ensure that the requirements of the GDC are satisfied. Imposition of ASME BPV Code Case N–770–2, with conditions, ensures that the requirements of the GDC are met for all mitigation techniques currently in use for Alloy 82/182 butt welds because ASME Code-allowable limits will not be exceeded, leakage would likely not occur and potential flaws will be detected before they challenge the structural or leak tight integrity of piping welds. All current licensees of U.S. pressurized water reactors will be required to implement ASME BPV Code Case N–770–2, as conditioned. The Code Case provisions on examination requirements for ASME Class 1 piping and nozzle nickel-alloy dissimilar metal butt welds are somewhat different from those established under ASME BPV Code Case N–770–1, as conditioned, and will require a licensee to modify its procedures for inspection of ASME Class 1 nickel-alloy welds to meet these requirements. Accordingly, the NRC imposition of the ASME BPV Code Case N–770–2, as conditioned, may be deemed to be a modification of the procedures to operate a facility resulting from the imposition of the new regulation, and as such, this rulemaking provision may be considered backfitting under § 50.109(a)(1). The NRC continues to find that ASME Class 1 nickel-alloy dissimilar metal weld inspections are necessary for adequate protection of public health and safety, and that the requirements of ASME BPV Code Case N–770–2, as conditioned, represent an acceptable approach developed by a voluntary consensus standards organization for performing future ASME Class 1 nickelalloy dissimilar metal weld inspections. The NRC concludes that approval of ASME BPV Code Case N–770–2, as conditioned, by incorporation by reference of the Code Case into § 50.55a, E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules is necessary to ensure that the facility provides adequate protection to the health and safety of the public and constitutes a redefinition of the requirements necessary to provide reasonable assurance of adequate protection of public health and safety. Therefore, a backfit analysis need not be prepared for this portion of the proposed rule in accordance with § 50.109(a)(4)(ii) and § 50.109(a)(4)(iii). Conclusion The NRC finds that incorporation by reference into § 50.55a of the 2009 Addenda through 2013 Edition of Section III, Division 1, of the ASME BPV Code subject to the identified conditions; the 2009 Addenda through 2013 Edition of Section XI, Division 1, of the ASME BPV Code, subject to the identified conditions; and the 2009 Edition through the 2012 Edition of the ASME OM Code subject to the identified conditions does not constitute backfitting or represent an inconsistency with any issue finality provisions in 10 CFR part 52. The NRC finds that the incorporation by reference of Code Cases N–824 and OMN–20 does not constitute backfitting or represent an inconsistency with any issue finality provisions in 10 CFR part 52. The NRC finds that the inclusion of a new condition on Code Case N–729–4 and a new condition on Code Case N– 770–2 constitutes backfitting necessary for adequate protection. XIV. Regulatory Flexibility Certification Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the NRC certifies that this proposed rule does not impose a significant economical impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of commercial nuclear power plants. A licensee who is a subsidiary of a large entity does not qualify as a small entity. The companies that own these plants 56855 are not ‘‘small entities’’ as defined in the Regulatory Flexibility Act or the size standards established by the NRC (10 CFR 2.810), as the companies: • Provide services that are not engaged in manufacturing, and have average gross receipts of more than $6.5 million over their last 3 completed fiscal years, and have more than 500 employees; • Are not governments of a city, county, town, township or village; • Are not school districts or special districts with populations of less than 50; and • Are not small educational institutions. XV. Availability of Documents The NRC is making the documents identified in Table 1 available to interested persons through one or more of the following methods, as indicated. To access documents related to this action, see the ADDRESSES section of this notice. TABLE 1—AVAILABILITY OF DOCUMENTS mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Document ADAMS Accession No. Proposed Rule Documents: Regulatory Analysis (includes backfitting discussion in Appendix A) ........................................................ Related Documents: Fatigue and Fracture Mechanics: 33rd Volume, ASTM STP 1417, W.G. Reuter and R.S. Piascik, Eds., ASTM International, West Conshohocken, PA, 2002. ‘‘Final Results from the CARINA Project on Crack Initiation and Arrest of Irradiated German RPV Steels for Neutron Fluences in the Upper Bound,’’ by AREVA at the 26th Symposium on Effects of Radiation on Nuclear Materials (June 12–13, 2013, Indianapolis, IN, USA). Letter from Brian Thomas, NRC, to Michael Merker, ASME; ‘‘Public Access to Material the NRC Seeks to Incorporate by Reference into its Regulations;’’ April 9, 2015. Letter from Michael Merker, ASME, to Brian Thomas, NRC; April 21, 2015 ............................................ Licensee Event Report 50–338/2012–001–00 ........................................................................................... NUREG–0800, ‘‘Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition’’. NUREG–0800, Section 3.9.6, Revision 3, ‘‘Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints,’’ March 2007. NUREG–0800, Section 5.4.2.2, Revision 1, ‘‘Steam Generator Tube Inservice Inspection,’’ July 1981 .. NUREG–1482, Revision 2, ‘‘Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants,’’ October 2013. NUREG–1801, Revision 2, ‘‘Generic Aging Lessons Learned (GALL) Report,’’ December 2010 ............ NUREG–1950, ‘‘Disposition of Public Comments and Technical Bases for Changes in the License Renewal Guidance Documents NUREG–1801 and NUREG–1800,’’ April 2011. NUREG/CR–6860, ‘‘An Assessment of Visual Testing,’’ November 2004 ................................................ NUREG/CR–6933, ‘‘Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods,’’ March 2007. NUREG/CR–7122, ‘‘An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,’’ March 2012. NRC Generic Letter 90–05, ‘‘Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90–05),’’ June 1990. NRC Meeting Summary of June 5–7, 2013, Annual Materials Programs Technical Information Exchange Public Meeting. NRC Memorandum, ‘‘Consolidation of SECY–94–084 and SECY–95–132,’’ July 24, 1995 .................... NRC Memorandum, ‘‘Staff Requirements—Affirmation Session, 11:30 a.m., Friday, September 10, 1999, Commissioners’ Conference Room, One White Flint North, Rockville, Maryland (Open to Public Attendance),’’ September 10, 1999. NRC Regulatory Guide 1.28, Revision 4, ‘‘Quality Assurance Program Criteria (Design and Construction),’’ June 2010. NRC Regulatory Guide 1.83, Revision 1, ‘‘Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes,’’ July 1975 (withdrawn in 2009). NRC Regulatory Guide 1.147, Revision 17, ‘‘Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,’’ August 2014. VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 PO 00000 Frm 00037 Fmt 4701 Sfmt 4702 E:\FR\FM\18SEP5.SGM ML14170B104. ML15085A206. ML15112A064. ML12151A441. ML070660036. ML070720041. ML052340627. ML13295A020. ML103490041. ML11116A062. ML043630040. ML071020410 and ML071020414. ML12087A004. ML031140590. ML14003A230. ML003708048. ML003755050. ML100160003. ML003740256. ML13339A689. 18SEP5 56856 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules TABLE 1—AVAILABILITY OF DOCUMENTS—Continued Document ADAMS Accession No. NRC Regulatory Guide 1.174, Revision 2, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,’’ May 2011. NRC Regulatory Guide 1.175, ‘‘An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing,’’ August 1998. NRC Regulatory Guide 1.192, Revision 1, ‘‘Operation and Maintenance Code Case Acceptability, ASME OM Code,’’ August 2014. NRC Regulatory Guide 1.200, Revision 2, ‘‘An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,’’ March 2009. NRC Regulatory Guide 1.201, Revision 1, ‘‘Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance,’’ May 2006. NRC Regulatory Information Conference, Recent Operating Reactors Materials Issues, Presentation Materials, 2013. Relief Request REP–1 U2, Revision 2 ....................................................................................................... SECY–90–016, ‘‘Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements’’. SECY–93–087, ‘‘Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs’’. SECY–94–084, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs’’. SECY–95–132, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY–94–084)’’. ASME Codes, Standards, and Code Cases: ASME BPV Code, Section III, Division 1: 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition. ASME BPV Code, Section XI, Division 1: 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition. ASME OM Code, Division 1: 2009 Edition, 2011 Addenda, and 2012 Edition ......................................... ASME Standard NQA–1: 1983 Edition through 1994 Edition, 2008 Edition, and 2009–1a Addenda ....... ASME BPV Code Case N–729–4 .............................................................................................................. ASME BPV Code Case N–770–2 .............................................................................................................. ASME BPV Code Case N–824 .................................................................................................................. ASME OM Code Case OMN–20 ................................................................................................................ Throughout the development of this rulemaking, the NRC may post documents related to this rule, including public comments, on the Federal rulemaking Web site at https:// www.regulations.gov under Docket ID NRC–2011–0088. The Federal rulemaking Web site allows you to receive alerts when changes or additions occur in a docket folder. To subscribe: (1) Navigate to the docket folder for NRC–2011–0088; (2) click the ‘‘Sign up for Email Alerts’’ link; and (3) enter your email address and select how frequently you would like to receive emails (daily, weekly, or monthly). mstockstill on DSK4VPTVN1PROD with PROPOSALS5 List of Subjects in 10 CFR Part 50 Administrative practice and procedure, Antitrust, Classified information, Criminal penalties, Education, Fire prevention, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalties, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements, Whistleblowing. For the reasons set forth in the preamble, and under the authority of the Atomic Energy Act of 1954, as amended; VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553, the NRC proposes to adopt the following amendments to 10 CFR part 50. PART 50—DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1. The authority citation for part 50 continues to read as follows: ■ Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Public Law 96–295, 94 Stat. 783. 2. In § 50.55a: a. Revise paragraphs (a) introductory text, (a)(1)(i) introductory text, (a)(1)(i)(E)(12), (a)(1)(i)(E)(13) and add paragraphs (a)(1)(i)(E)(14) through (a)(1)(i)(E)(17); ■ b. Revise paragraph (a)(1)(ii) introductory text, (a)(1)(ii)(C)(48) and ■ ■ PO 00000 Frm 00038 Fmt 4701 Sfmt 4702 ML100910006. ML003740149. ML13340A034. ML090410014. ML061090627. https://www.nrc.gov/public-involve/ conference-symposia/ric/past/ 2013/docs/abstracts/ sessionabstract-19.html. ML13232A308. ML003707849. ML003708021. ML003708068. ML003708005. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. https://go.asme.org/NRC. (a)(1)(ii)(C)(49) and add paragraphs (a)(1)(ii)(C)(50) through (a)(1)(ii)(C)(53); ■ c. Revise paragraphs (a)(1)(iii)(B) and (a)(1)(iii)(C) and add paragraphs (a)(1)(iii)(D), (a)(1)(iii)(E); ■ d. Revise paragraphs (a)(1)(iv) introductory text and add paragraphs (a)(1)(iv)(B) and (a)(1)(iv)(C); ■ e. Add paragraph (a)(1)(v); ■ f. Revise paragraphs (b) introductory text, (b)(1) introductory text, (b)(1)(ii), (b)(1)(iv), and (b)(1)(vii) and add paragraph (b)(1)(viii); ■ g. Revise paragraphs (b)(2) introductory text, (b)(2)(vi); ■ h. Revise paragraph (b)(2)(viii) introductory text and add paragraphs (b)(2)(viii)(H) and (b)(2)(viii)(I); ■ i. Revise paragraphs (b)(2)(ix) introductory text, (b)(2)(ix)(D), (b)(2)(x), add paragraph (b)(2)(xviii)(D), revise paragraph (b)(2)(xxi)(A), and add paragraphs (b)(2)(xxx) through (b)(2)(xxxvii); ■ j. Revise paragraphs (b)(3) introductory text, (b)(3)(i), and (b)(3)(ii), add paragraph (b)(3)(iii), revise paragraphs (b)(3)(iv) introductory text and (b)(3)(iv)(A) though (b)(3)(iv)(D), and add paragraphs (b)(3)(vii) through (b)(3)(xi); E:\FR\FM\18SEP5.SGM 18SEP5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules k. Revise paragraphs (b)(4) introductory text, (b)(5), and (b)(6); ■ l. Revise paragraphs (f) introductory text, (f)(2), (f)(3)(iii)(A), (f)(3)(iii)(B), (f)(3)(iv)(A), (f)(3)(iv)(B), (f)(4) introductory text, (f)(4)(i), (f)(4)(ii); ■ m. Revise paragraphs (g) introductory text, (g)(2), (g)(3) introductory text, (g)(3)(i), (g)(3)(ii), (g)(3)(v), (g)(4)(i), (g)(4)(ii), and (g)(6)(ii)(D)(1) through (g)(6)(ii)(D)(4), remove paragraphs (g)(6)(ii)(D)(5) and (g)(6)(ii)(D)(6), revise paragraphs (g)(6)(ii)(F)(1) through (g)(6)(ii)(F)(10), and add paragraphs (g)(6)(ii)(F)(11) through (g)(6)(ii)(F)(13). The revisions and additions read as follows: ■ mstockstill on DSK4VPTVN1PROD with PROPOSALS5 § 50.55a Codes and standards. (a) Documents approved for incorporation by reference. The standards listed in this paragraph have been approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. The standards are available for inspection, by appointment, at the NRC Technical Library, which is located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301–415–7000; email: Library.Resource@nrc.gov; or at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202–741–6030 or go to https://www.archives.gov/federalregister/cfr/ibr-locations.html. (1) * * * (i) ASME Boiler and Pressure Vessel Code, Section III. The editions and addenda for Section III of the ASME Boiler and Pressure Vessel Code (excluding Nonmandatory Appendices) are listed below, but limited by those provisions identified in paragraph (b)(1) of this section. * * * * * (E) * * * (12) 2007 Edition, (13) 2008 Addenda, (14) 2009 Addenda, (15) 2010 Edition, (16) 2011 Addenda, and (17) 2013 Edition. (ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions and addenda for Section XI of the ASME Boiler and Pressure Vessel Code (excluding Nonmandatory Appendix U) are listed below, but limited by those VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 provisions identified in paragraph (b)(2) of this section. * * * * * (C) * * * (48) 2007 Edition, (49) 2008 Addenda, (50) 2009 Addenda, (51) 2010 Edition, (52) 2011 Addenda, and (53) 2013 Edition. (iii) * * * (B) ASME BPV Code Case N–729–4. ASME BPV Code Case N–729–4, ‘‘Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI, Division 1’’ (Approval Date: June 22, 2012), with the conditions in paragraph (g)(6)(ii)(D) of this section. (C) ASME BPV Code Case N–770–2. ASME BPV Code Case N–770–2, ‘‘Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities Section XI, Division 1’’ (Approval Date: June 9, 2011), with the conditions in paragraph (g)(6)(ii)(F) of this section. (D) ASME BPV Code Case N–824. ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1’’ (Approval Date: October 16, 2012), with the conditions in paragraphs (b)(2)(xxxvii)(A) through (E) of this section. (E) ASME OM Code Case OMN–20. ASME OM Code Case OMN–20, ‘‘Inservice Test Frequency,’’ in the 2012 Edition of the ASME OM Code. OMN– 20 is referenced in paragraph (b)(3)(x). (iv) ASME Operation and Maintenance Code. The editions and addenda for the ASME Operation and Maintenance of Nuclear Power Plants are listed below, but limited by those provisions identified in paragraph (b)(3) of this section. * * * * * (B) ‘‘Operation and Maintenance of Nuclear Power Plants, Division 1: Section IST Rules for Inservice Testing of Light-Water Reactor Power Plants’’ (1) 2009 Edition and (2) 2011 Addenda. (C) ‘‘Operation and Maintenance of Nuclear Power Plants, Division 1: OM Code: Section IST.’’ PO 00000 Frm 00039 Fmt 4701 Sfmt 4702 56857 (1) 2012 Edition. (2) [Reserved] (v) ASME Quality Assurance Requirements. (A) ASME NQA–1, ‘‘Quality Assurance Program Requirements for Nuclear Facilities.’’ (1) NQA–1–1983 Edition, (2) NQA–1a–1983 Addenda, (3) NQA–1b–1984 Addenda, (4) NQA–1c–1985 Addenda, (5) NQA–1–1986 Edition, (6) NQA–1a–1986 Addenda, (7) NQA–1b–1987 Addenda, and (8) NQA–1c–1988 Addenda. (9) NQA–1–1989 Edition, (10) NQA–1a–1989 Addenda, (11) NQA–1b–1991 Addenda, and (12) NQA–1c–1992 Addenda. (B) ASME NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications.’’ (1) NQA–1–1994 Edition, (2) NQA–1a–2008 Edition, and (3) NQA–1a–2009 Addenda. * * * * * (b) Use and conditions on the use of standards. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code (BPV Code) and the ASME Operation and Maintenance of Nuclear Power Plants (OM Code) as specified in this paragraph. Each combined license for a utilization facility is subject to the following conditions. (1) Conditions on ASME BPV Code Section III. Each manufacturing license, standard design approval, and design certification under part 52 of this chapter is subject to the following conditions. As used in this section, references to Section III refer to Section III of the ASME Boiler and Pressure Vessel Code and include the 1963 Edition through 1973 Winter Addenda and the 1974 Edition (Division 1) through the 2013 Edition (Division 1), subject to the following conditions: * * * * * (ii) Section III condition: Weld leg dimensions. When applying the 1989 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1) of this section, applicants and licensees may not apply the Section III provisions identified in Table 1 of this section for welds with leg size less than 1.09 tn E:\FR\FM\18SEP5.SGM 18SEP5 56858 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules TABLE 1 OF § 50.55A—PROHIBITED CODE PROVISIONS Editions and Addenda Code provision 1989 Addenda through 2013 Edition ................................................................................................................ Subparagraph NB–3683.4(c)(1). Subparagraph NB–3683.4(c)(2). Note 11 to Figure NC–3673.2(b)–1. Note 11 to Figure ND–3673.2(b)–1. Note 13 to Figure NC–3673.2(b)–1. Note 13 to Figure ND–3673.2(b)–1. Note 11 to Table NC–3673.2(b)–1. Note 11 to Table ND–3673.2(b)–1. 1989 Addenda through 2003 Addenda ............................................................................................................. 2004 Edition through 2010 Edition ................................................................................................................... 2011 Addenda through 2013 Edition ................................................................................................................ mstockstill on DSK4VPTVN1PROD with PROPOSALS5 * * * * * (iv) Section III condition: Quality assurance. When applying editions and addenda later than the 1989 Edition of Section III, the requirements of NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications,’’ 1983 Edition through the 1994 Edition, 2008 Edition, and the 2009–1a Addenda specified in either NCA–4000 or NCA– 7000 of that edition and addenda of Section III may be used by an applicant or licensee provided that the administrative, quality, and technical provisions contained in that edition and addenda of Section III are used in conjunction with the applicant’s or licensee’s appendix B to 10 CFR part 50 quality assurance program; and that commitments contained in the applicant’s or licensee’s quality assurance program description which are either more stringent than those contained in NQA–1 or have no comparable provision in NQA–1 or Section III, govern the applicant’s or licensee’s Section III activities. * * * * * (vii) Section III condition: Capacity certification and demonstration of function of incompressible-fluid pressure-relief valves. When applying the 2006 Addenda through the 2013 Edition, applicants and licensees may use paragraph NB–7742, except that paragraph NB–7742(a)(2) may not be used. For a valve design of a single size to be certified over a range of set pressures, the demonstration of function tests under paragraph NB–7742 must be conducted as prescribed in NB–7732.2 on two valves covering the minimum set pressure for the design and the maximum set pressure that can be accommodated at the demonstration facility selected for the test. (viii) Section III condition: Use of ASME certification marks. When applying editions and addenda earlier than the 2011 Addenda to the 2010 Edition, licensees may use either the ASME BPV Code Symbol Stamps or the ASME Certification Marks with the appropriate certification designators and class designators as specified in the VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 2013 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1) of this section. (2) Conditions on ASME BPV Code, Section XI. As used in this section, references to Section XI refer to Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition through the 1976 Winter Addenda and the 1977 Edition through the 2013 Edition (excluding Nonmandatory Appendix U), subject to the following conditions: * * * * * (vi) Section XI condition: Effective edition and addenda of Subsection IWE and Subsection IWL. Licensees that implemented the expedited examination of containment, in accordance with Subsection IWE and Subsection IWL, during the period from September 9, 1996, to September 9, 2001, may use either the 1992 Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned by the requirements in paragraphs (b)(2)(viii) and (ix) of this section, when implementing the initial 120-month inspection interval for the containment inservice inspection requirements of this section. Successive 120-month interval updates must be implemented in accordance with paragraph (g)(4)(ii) of this section. * * * * * (viii) Section XI condition: Concrete containment examinations. Applicants or licensees applying Subsection IWL, 1992 Edition with the 1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through (E) of this section. Applicants or licensees applying Subsection IWL, 1995 Edition with the 1996 Addenda, must apply paragraphs (b)(2)(viii)(A), (b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of this section. Applicants or licensees applying Subsection IWL, 1998 Edition through the 2000 Addenda, must apply paragraphs (b)(2)(viii)(E) and (F) of this section. Applicants or licensees applying Subsection IWL, 2001 Edition through the 2004 Edition, up to and including the 2006 Addenda, must apply paragraphs (b)(2)(viii)(E) through PO 00000 Frm 00040 Fmt 4701 Sfmt 4702 (G) of this section. Applicants or licensees applying Subsection IWL, 2007 Edition up to and including the 2008 Addenda must apply paragraph (b)(2)(viii)(E) of this section. Applicants or licensees applying Subsection IWL, 2007 Edition with the 2009 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must apply paragraph (b)(2)(viii)(H) and (b)(2)(viii)(I) of this section. * * * * * (H) Concrete containment examinations: Eighth provision. For each inaccessible area of concrete identified for evaluation under IWL– 2512, the licensee must provide the applicable information specified in paragraphs (b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and (b)(2)(viii)(E)(3) of this section in the ISI Summary Report required by IWA–6000. (I) Concrete containment examinations: Ninth provision. During the period of extended operation of a renewed license under part 54 of this chapter, the licensee must perform the technical evaluation under IWL–2512(b) of inaccessible below-grade concrete surfaces exposed to foundation soil, backfill, or groundwater at periodic intervals not to exceed 5 years. In addition, the licensee must examine representative samples of the exposed portions of the below-grade concrete, when such below-grade concrete is excavated for any reason. * * * * * (ix) Section XI condition: Metal containment examinations. Applicants or licensees applying Subsection IWE, 1992 Edition with the 1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A) through (E) of this section. Applicants or licensees applying Subsection IWE, 1998 Edition through the 2001 Edition with the 2003 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) through (I) of this section. Applicants or licensees applying Subsection IWE, 2004 Edition, up to and including the 2005 Addenda, must E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) through (H) of this section. Applicants or licensees applying Subsection IWE, 2004 Edition with the 2006 Addenda, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) of this section. Applicants or licensees applying Subsection IWE, 2007 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) of this section. * * * * * (D) Metal containment examinations: Fourth provision. This paragraph (b)(2)(ix)(D) may be used as an alternative to the requirements of IWE– 2430. If the examinations reveal flaws or areas of degradation exceeding the acceptance standards of Table IWE– 3410–1, an evaluation must be performed to determine whether additional component examinations are required. For each flaw or area of degradation identified that exceeds acceptance standards, the applicant or licensee must provide the following in the ISI Summary Report required by IWA–6000: (1) A description of each flaw or area, including the extent of degradation, and the conditions that led to the degradation; (2) The acceptability of each flaw or area and the need for additional examinations to verify that similar degradation does not exist in similar components; (3) A description of necessary corrective actions; and (4) The number and type of additional examinations to ensure detection of similar degradation in similar components. * * * * * (x) Section XI condition: Quality assurance. When applying the editions and addenda later than the 1989 Edition of ASME BPV Code, Section XI, the edition and addenda of NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications,’’ 1983 Edition through the 1994 Edition, the 2008 Edition, and the 2009–1a Addenda specified in either IWA–1400 or Table IWA 1600–1 of that edition and addenda of Section XI, may be used by a licensee provided that the licensee uses its appendix B to 10 CFR part 50 quality assurance program in conjunction with Section XI requirements. Commitments contained in the licensee’s quality assurance program description that are more stringent than those contained in NQA– VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 1 must govern Section XI activities. Further, where NQA–1 and Section XI do not address the commitments contained in the licensee’s appendix B quality assurance program description, the commitments must be applied to Section XI activities. * * * * * (xviii) * * * (D) NDE personnel certification: Fourth provision. The use of Appendix VII and subarticle VIII–2200 of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV Code is prohibited. When using ASME BPV Code, Section XI editions and addenda later than the 2010 Edition, licensees and applicants must use the prerequisites for ultrasonic examination personnel certifications in Table VII–4110–1 and subarticle VIII– 2200, Appendix VIII in the 2010 Edition. * * * * * (xxi) * * * (A) Table IWB–2500–1 examination requirements: First provision. The provisions of Table IWB 2500–1, Examination Category B–D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B) of the 1998 Edition must be applied when using the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section. A visual examination with magnification that has a resolution sensitivity to resolve 0.044 inch (1.1 mm) lower case characters without an ascender or descender (e.g., a, e, n, v), utilizing the allowable flaw length criteria in Table IWB–3512–1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed instead of an ultrasonic examination. * * * * * (xxx) Section XI condition: Steam generator preservice examinations. Prior to plant start up with a newly installed steam generator, a 100 percent full length examination will be conducted of the tubing in each new steam generator instead of the preservice inspection requirements of IWB–2200(c). (xxxi) Section XI condition: Mechanical clamping devices. The use of mechanical clamping devices on Class 1 piping and portions of piping systems that form the containment boundary is prohibited. (xxxii) Section XI condition: Summary report submittal. When using ASME BPV Code, Section XI, 2010 Edition through the latest edition and PO 00000 Frm 00041 Fmt 4701 Sfmt 4702 56859 addenda incorporated by reference in paragraph (a)(1)(ii) of this section, Summary Reports described in IWA– 6000 must be submitted to the NRC. Preservice inspection summary reports shall be submitted prior to the date of placement of the unit into commercial service and inservice inspection summary reports shall be submitted within 90 calendar days of the completion of each refueling outage. (xxxiii) Section XI condition: RiskInformed allowable pressure. The use of Paragraph G–2216 in Appendix G in the 2011 Addenda and later editions and addenda of the ASME BPV Code, Section XI is prohibited. (xxxiv) Section XI condition: Disposition of flaws in Class 3 components. When using the 2013 Edition of the ASME BPV Code, Section XI, to disposition flaws in Examination Category D–A components (i.e., welded attachments for vessels, piping, pumps, and valves), the acceptance standards of IWD–3510 must be used. (xxxv) Section XI condition: Use of RTT0 in the KIa and KIc equations. When using the 2013 Edition of the ASME BPV Code, Section XI, Appendix A, paragraph A–4200, if T0 is available, then RTT0 may be used in place of RTNDT for applications using the KIc equation and the associated KIc curve, but not for applications using the KIa equation and the associated KIa curve. (xxxvi) Section XI condition: Fracture toughness of irradiated materials. When using the 2013 Edition of the ASME BPV Code, Section XI, Appendix A paragraph A–4400, the licensee shall obtain NRC approval before using irradiated T0 and the associated RTT0 in establishing fracture toughness of irradiated materials. (xxxvii) Section XI condition: ASME BPV Code Case N–824. Licensees may use the provisions of ASME BPV Code Case N–824, ‘‘Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface Section XI, Division 1,’’ subject to the following conditions. (A) Ultrasonic examinations must be spatially encoded. (B) Instead of Paragraph 1(c)(1)(–a) licensees shall use dual, transmitreceive, refracted longitudinal wave, multi-element phased array search units. (C) Instead of Paragraph 1(c)(1)(–c) (–1), licensees shall use a phased array search unit with a center frequency between 500 kHz and 1 MHz. (D) Instead of Paragraph 1(c)(1)(–c) (–2), licensees shall use a phased array search unit with a center frequency of 500 kHz. (E) Instead of Paragraph 1(c)(1)(–d), the phased array search unit must E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56860 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules produce angles from 30 to 70 degrees with a maximum increment of 5 degrees. (3) Conditions on ASME OM Code. As used in this section, references to the OM Code are to the ASME OM Code, Subsections ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and Nonmandatory Appendices A through H and J through M, in the 1995 Edition through the 2012 Edition as specified in paragraph (a)(1)(iv). The following conditions are applicable when implementing the ASME OM Code: (i) OM condition: Quality assurance. When applying editions and addenda of the OM Code, the requirements of ASME Standard NQA–1, ‘‘Quality Assurance Requirements for Nuclear Facility Applications,’’ 1983 Edition through the 1994 Edition, 2008 Edition, and 2009–1a Addenda, are acceptable as permitted by either ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA–1500 of the 1998 Edition through the latest edition and addenda of the OM Code incorporated by reference in paragraph (a)(1)(iv) of this section, provided the licensee uses its appendix B to 10 CFR part 50 quality assurance program in conjunction with the OM Code requirements. Commitments contained in the licensee’s quality assurance program description that are more stringent than those contained in NQA–1 govern OM Code activities. If NQA–1 and the OM Code do not address the commitments contained in the licensee’s appendix B quality assurance program description, the commitments must be applied to OM Code activities. (ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees must comply with the provisions for testing MOVs in OM Code, ISTC 4.2, 1995 Edition with the 1996 and 1997 Addenda, or ISTC–3500, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, and must establish a program to ensure that MOVs continue to be capable of performing their design basis safety functions. Licensees implementing OM Code, Mandatory Appendix III, ‘‘Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,’’ of the 2009 Edition, 2011 Addenda, and 2012 Edition shall comply with the following conditions: (A) MOV diagnostic test interval. Licensees shall evaluate the adequacy of the diagnostic test interval for each MOV and adjust the interval as necessary, but not later than 5 years or three refueling outages (whichever is VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 longer) from initial implementation of OM Code, Appendix III. (B) MOV testing impact on risk. Licensees shall ensure that the potential increase in core damage frequency and large early release frequency associated with the extension is acceptably small when extending exercise test intervals for high risk MOVs beyond a quarterly frequency. (C) MOV risk categorization. When applying Appendix III to the OM Code, licensees shall categorize MOVs according to their safety significance using the methodology described in ASME OM Code Case OMN–3, ‘‘Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power Plants,’’ subject to the conditions applicable to OMN–3 which are set forth in Regulatory Guide 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety evaluation. (D) MOV stroke time. When applying Paragraph III–3600, ‘‘MOV Exercising Requirements,’’ of Appendix III to the OM Code, licensees shall verify that the stroke time of the MOV satisfies the assumptions in the plant safety analyses. (iii) OM condition: New Reactors. In addition to complying with the provisions in the OM Code with the conditions specified in paragraph (b)(3) of this section, holders of operating licenses for nuclear power reactors that received construction permits under this part on or after the date 12 months after [the effective date of the final rule], and holders of combined licenses issued under 10 CFR part 52, whose initial fuel loading occurs on or after the date 12 months after [the effective date of the final rule] shall also comply with the following conditions, as applicable: (A) Power-operated valves. Licensees shall periodically verify the capability of power-operated valves to perform their design-basis safety functions. (B) Check valves. Licensees must perform bi-directional testing of check valves within the IST program where practicable. (C) Flow-induced vibration. Licensees shall monitor flow-induced vibration from hydrodynamic loads and acoustic resonance during preservice testing and inservice testing to identify potential adverse flow effects on components within the scope of the IST program. (D) High risk non-safety systems. Licensees shall assess the operational readiness of pumps, valves, and dynamic restraints within the scope of PO 00000 Frm 00042 Fmt 4701 Sfmt 4702 the Regulatory Treatment of Non-Safety Systems for applicable reactor designs. (iv) OM condition: Check valves (Appendix II). Appendix II, ‘‘Check Valve Condition Monitoring Program,’’ of the OM Code, 2003 Addenda through the 2012 Edition, is acceptable for use without conditions with the clarifications that (1) the maximum test interval allowed by Appendix II for individual check valves in a group of two valves or more must be supported by periodic testing of a sample of check valves in the group during the allowed interval and (2) the periodic testing plan must be designed to test each valve of a group at approximate equal intervals not to exceed the maximum requirement interval. Licensees applying Appendix II of the OM Code, 1995 Edition with the 1996 and 1997 Addenda, shall satisfy the requirements of paragraphs (b)(3)(iv)(A) through (C) of this section. Licensees applying Appendix II, 1998 Edition through the 2012 Edition, shall satisfy the requirements of paragraphs (b)(3)(iv)(A), (B), and (D) of this section. * * * * * (vii) OM condition: Subsection ISTB. Subsection ISTB, 2011 Addenda, is prohibited for use. (viii) OM condition: Subsection ISTE. Licensees may not implement the riskinformed approach for inservice testing (IST) of pumps and valves specified in Subsection ISTE, ‘‘Risk-Informed Inservice Testing of Components in Light-Water Reactor Nuclear Power Plants,’’ in the OM Code, 2009 Edition, 2011 Addenda, or 2012 Edition, without first obtaining NRC authorization to use Subsection ISTE as an alternative to the applicable IST requirements in the OM Code pursuant to § 50.55a(z). (ix) OM condition: Subsection ISTF. Licensees applying Subsection ISTF, 2012 Edition, shall satisfy the requirements of Mandatory Appendix V, ‘‘Pump Periodic Verification Test Program,’’ of the ASME OM Code, 2012 Edition. Subsection ISTF, 2011 Addenda, is not acceptable for use. (x) OM condition: ASME OM Code Case OMN–20. Licensees may implement ASME OM Code Case OMN– 20, ‘‘Inservice Test Frequency,’’ which is incorporated by reference in paragraph (a)(1)(iii)(E) of this section. (xi) OM condition: Valve Position Indication. When implementing ASME OM Code, Subsection ISTC–3700, ‘‘Position Verification Testing,’’ licensees shall develop and implement a method to verify that valve operation is accurately indicated by supplementing valve position indicating lights with other indications, such as flow meters or other suitable E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules instrumentation, to provide assurance of proper obturator position. (4) Conditions on Design, Fabrication, and Materials Code Cases. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter is subject to the following conditions. Licensees may apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.84, as incorporated by reference in paragraph (a)(3)(i) of this section, without prior NRC approval, subject to the following conditions: * * * * * (5) Conditions on inservice inspection Code Cases. Licensees may apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as incorporate by reference in paragraph (a)(3)(ii) of this section, without prior NRC approval, subject to the following: (i) ISI Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section. (ii) ISI Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month ISI program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section. (iii) ISI Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in NRC Regulatory Guide 1.147. If a licensee has applied a listed Code Case that is later listed as annulled in NRC Regulatory Guide 1.147, the licensee may continue to apply the Code Case to the end of the current 120-month interval. (6) Conditions on Operation and Maintenance of Nuclear Power Plants Code Cases. Licensees may apply the ASME Operation and Maintenance Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by reference in VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 paragraph (a)(3)(iii), without prior NRC approval, subject to the following: (i) OM Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section. (ii) OM Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month ISI program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of NRC Regulatory Guide 1.192, as incorporated by reference in paragraph (a)(3)(iii) of this section. (iii) OM Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in NRC Regulatory Guide 1.192. If a licensee has applied a listed Code Case that is later listed as annulled in NRC Regulatory Guide 1.192, the licensee may continue to apply the Code Case to the end of the current 120-month interval. * * * * * (f) Inservice testing requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph collectively as inservice testing) of the ASME BPV Code and ASME OM Code as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions. Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (f)(4) through (6) of this section must be met only after the Commission makes the finding under § 52.103(g) of this chapter. Requirements for inservice inspection of Class 1, Class 2, Class 3, Class MC, and Class CC components (including their supports) are located in § 50.55a(g). * * * * * PO 00000 Frm 00043 Fmt 4701 Sfmt 4702 56861 (2) Design and accessibility requirements for performing inservice testing in plants with CPs issued between 1971 and 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, pumps and valves that are classified as ASME Code Class 1 and Class 2 must be designed and provided with access to enable the performance of inservice tests for operational readiness set forth in editions and addenda of Section XI of the ASME BPV incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 1.192, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) in effect 6 months before the date of issuance of the construction permit. The pumps and valves may meet the inservice test requirements set forth in subsequent editions of this Code and addenda that are incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 1.192, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the applicable conditions listed therein. * * * * * (3) * * * (iii) * * * (A) Class 1 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 1 must be designed and provided with access to enable the performance of inservice testing of those pumps and valves within the scope of the ASME OM Code for assessing operational readiness, as set forth in either the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 1.192, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) which are applied to the construction of the particular pump or valve or the summer 1973 Addenda, whichever is later. (B) Class 1 pumps and valves: Second provision. In facilities whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, issued on or after November 22, 1999, E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 56862 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules pumps and valves that are classified as ASME Code Class 1 must be designed and provided with access to enable the performance of inservice testing of those pumps and valves within the scope of the ASME OM Code for assessing operational readiness, as set forth in editions and addenda of the ASME OM Code (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by reference in paragraph (a)(3)(iii) of this section), incorporated by reference in paragraph (a)(1)(iv) of this section at the time the construction permit, combined license, manufacturing license, design certification, or design approval is issued. (iv) * * * (A) Class 2 and 3 pumps and valves: First provision. In facilities whose construction permit was issued before November 22, 1999, pumps and valves that are classified as ASME Code Class 2 and Class 3 that are within the scope of the ASME OM Code and are not covered by paragraph (f)(3)(iii)(A) of this section must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in the editions and addenda of Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later. (B) Class 2 and 3 pumps and valves: Second provision. In facilities whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, issued on or after November 22, 1999, pumps and valves that are classified as ASME Code Class 2 and 3 that are within the scope of the ASME OM Code and are not covered by paragraph (f)(3)(iii)(B) of this section must be designed and provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in editions and addenda of the ASME OM Code (or the optional ASME OM Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by reference in paragraph (a)(3)(iii) of this section), incorporated by reference in paragraph (a)(1)(iv) of this section at the time the construction permit, combined license, or design certification is issued. * * * * * VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 (4) Inservice testing standards requirement for operating plants. Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. (i) Applicable IST Code: Initial 120month interval. Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during the initial 120month interval must comply with the requirements in the latest edition and addenda of the OM Code incorporated by reference in paragraph (a)(1)(iv) of this section on the date 12 months before the date of issuance of the operating license under this part, or 12 months before the date scheduled for initial loading of fuel under a combined license under part 52 of this chapter (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by reference in paragraph (a)(3)(iii) of this section, subject to the conditions listed in paragraph (b) of this section). (ii) Applicable IST Code: Successive 120-month intervals. Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 12 months before the start of the 120-month interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 1.192 as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the conditions listed in paragraph (b) of this section. * * * * * (g) Preservice and inservice inspection requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions. Each combined license for a PO 00000 Frm 00044 Fmt 4701 Sfmt 4702 boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (g)(4) through (6) of this section must be met only after the Commission makes the finding under § 52.103(g) of this chapter. Requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in § 50.55a(f). * * * * * (2) Accessibility requirements—(i) Accessibility requirements for plants with CPs issued between 1971 and 1974. For a boiling or pressurized watercooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, components that are classified as ASME Code Class 1 and Class 2 and supports for components that are classified as ASME Code Class 1 and Class 2 must be designed and be provided with the access necessary to perform the required preservice and inservice examinations set forth in editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section) in effect 6 months before the date of issuance of the construction permit. (ii) Accessibility requirements for plants with CPs issued after 1974. For a boiling or pressurized water-cooled nuclear power facility, whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, was issued on or after July 1, 1974, components that are classified as ASME Code Class 1, Class 2, and Class 3 and supports for components that are classified as ASME Code Class 1, Class 2, and Class 3 must be designed and provided with the access necessary to perform the required preservice and inservice examinations set forth in editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular component. (iii) Accessibility requirements: Meeting later Code requirements. All components (including supports) may meet the requirements set forth in subsequent editions of codes and addenda or portions thereof that are E:\FR\FM\18SEP5.SGM 18SEP5 mstockstill on DSK4VPTVN1PROD with PROPOSALS5 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules incorporated by reference in paragraph (a) of this section, subject to the conditions listed therein. (3) Preservice examination requirements—(i) Preservice examination requirements for plants with CPs issued between 1971 and 1974. For a boiling or pressurized watercooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, components that are classified as ASME Code Class 1 and Class 2 and supports for components that are classified as ASME Code Class 1 and Class 2 must meet the preservice examination requirements set forth in editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section) in effect 6 months before the date of issuance of the construction permit. (ii) Preservice examination requirements for plants with CPs issued after 1974. For a boiling or pressurized water-cooled nuclear power facility, whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, was issued on or after July 1, 1974, components that are classified as ASME Code Class 1, Class 2, and Class 3 and supports for components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the preservice examination requirements set forth in the editions and addenda of Section III or Section XI of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular component. * * * * * (v) Preservice examination requirements: Meeting later Code requirements. All components (including supports) may meet the requirements set forth in subsequent editions of codes and addenda or portions thereof that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed therein. * * * * * (4) * * * (i) Applicable ISI Code: Initial 120month interval. Inservice examination of components and system pressure VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 tests conducted during the initial 120month inspection interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section on the date 12 months before the date of issuance of the operating license under this part, or 12 months before the date scheduled for initial loading of fuel under a combined license under part 52 of this chapter (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, when using Section XI, or NRC Regulatory Guide 1.192, when using the OM Code, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the conditions listed in paragraph (b) of this section. (ii) Applicable ISI Code: Successive 120-month intervals. Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, when using Section XI, or NRC Regulatory Guide 1.192, when using the OM Code, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section. However, a licensee whose inservice inspection interval commences during the 12 through 18-month period after July 21, 2011, may delay the update of their Appendix VIII program by up to 18 months after July 21, 2011. * * * * * (6) * * * (ii) * * * (D) * * * (1) Implementation: Holders of operating licenses or combined licenses for pressurized-water reactors as of or after [the effective date of the final rule] shall implement the requirements of ASME BPV Code Case N–729–4 instead of ASME BPV Code Case N–729–1, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (4) of this section, by the first refueling outage starting after [the effective date of the final rule]. (2) Appendix I use: Appendix I of ASME BPV Code Case N–729–4 shall not be implemented without prior NRC approval. (3) Bare metal visual frequency: Instead of Note 4 of ASME BPV Code Case N–729–4, the following shall be PO 00000 Frm 00045 Fmt 4701 Sfmt 4702 56863 implemented; If EDY<8 and if no flaws are found that are attributed to PWSCC; (a) A bare metal visual examination is not required during refueling outages when a volumetric or surface examination is performed, (b) If a wetted surface examination has been performed of all of the partial penetration welds during the previous non-visual examination, the reexamination frequency may be extended to every third refueling outage or 5 calendar years, whichever is less, provided an IWA–2212 VT–2 visual examination of the head is performed under the insulation through multiple access points in outages that the VE is not completed. This IWA–2212 VT–2 visual examination may be performed with the reactor vessel depressurized. (4) Surface exam acceptance criteria: In addition to the requirements of paragraph-3132.1(b) of ASME BPV Code Case N–729–4, a component whose surface examination detects rounded indications greater than allowed in Paragraph NB–5352 in size on the partial-penetration or associated fillet weld shall be classified as having an unacceptable indication and corrected in accordance with the provisions of paragraph-3132.2 of ASME BPV Code Case N–729–4. * * * * * (F) * * * (1) Implementation: Holders of operating licenses or combined licenses for pressurized-water reactors as of or after [the effective date of the final rule] shall implement the requirements of ASME BPV Code Case N–770–2 instead of ASME BPV Code Case N–770–1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of this section, by the first refueling outage starting after [the effective date of the final rule]. (2) Categorization: Full structural weld overlays, authorized by the NRC staff in accordance with the alternatives approval process of this section, may be categorized as Inspection Items C–1 or F–1, as appropriate. Welds that have been mitigated by the Mechanical Stress Improvement Process (MSIPTM) may be categorized as Inspection Items D or E, as appropriate, provided the criteria in Appendix I of the code case have been met. For the purpose of determining ISI frequencies, all other butt welds that rely on Alloy 82/182 for structural integrity shall be categorized as Inspection Items A–1, A–2, or B until the NRC staff has reviewed the mitigation and authorized an alternative code case Inspection Item for the mitigated weld, or an alternative code case Inspection Item is used based on E:\FR\FM\18SEP5.SGM 18SEP5 56864 Federal Register / Vol. 80, No. 181 / Friday, September 18, 2015 / Proposed Rules mstockstill on DSK4VPTVN1PROD with PROPOSALS5 conformance with an ASME mitigation code case endorsed in NRC Regulatory Guide 1.147 with any applying conditions specified in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section. Paragraph-1100(e) of ASME BPV Code Case N–770–2 shall not be used to exempt welds that rely on Alloy 82/182 for structural integrity from any requirement of paragraph (g)(6)(ii)(F) of this section. (3) Baseline examinations: Baseline examinations for welds in Table 1 of ASME BPV Code Case N–770–2, Inspection Items A–1, A–2, and B, if not previously performed or currently scheduled to be performed in an ongoing refueling outage at the time this rule becomes effective, in accordance with paragraph (g)(6)(ii)(F) of this section, shall be completed by the end of the next refueling outage. Previous examinations of these welds can be credited for baseline examinations only if they were performed within the reinspection period for the weld item in Table 1 of ASME BPV Code Case N– 770–2 and the examination of each weld meets the examination requirements of paragraphs -2500(a) or -2500(b) of ASME BPV Code Case N–770–2. Other previous examinations that do not meet these requirements can be used to meet the baseline examination requirement, provided NRC approval in accordance with paragraphs (z)(1) or (2) of this section, is granted prior to the end of the next refueling outage. (4) Examination coverage: When implementing paragraph-2500(a) of ASME Code Case N–770–2, essentially 100 percent volumetric examination coverage shall be obtained, including greater than 90 percent volumetric examination coverage for circumferential flaws. Licensees are VerDate Sep<11>2014 21:45 Sep 17, 2015 Jkt 235001 prohibited from using Paragraph-2500(c) and -2500(d) of ASME BPV Code Case N–770–2 to meet examination requirements. (5) Inlay/onlay inspection frequency: All hot-leg operating temperature welds in Inspection Items G, H, J, and K shall be inspected each inspection interval. A 25 percent sample of Inspection Items G, H, J, and K cold-leg operating temperature welds shall be inspected whenever the core barrel is removed (unless it has already been inspected within the past 10 years) or within 20 years, whichever is less. (6) Reporting requirements: For any mitigated weld whose volumetric examination detects growth of existing flaws in the required examination volume that exceed the previous IWB– 3600 flaw evaluations or new flaws, a report summarizing the evaluation, along with inputs, methodologies, assumptions, and causes of the new flaw or flaw growth is to be provided to the NRC prior to the weld being placed in service other than modes 5 or 6. (7) Defining ‘‘t’’: For Inspection Items G, H, J, and K, when applying the acceptance standards of ASME BPV Code, Section XI, IWB–3514, for planar flaws contained within the inlay or onlay, the thickness ‘‘t’’ in IWB–3514 is the thickness of the inlay or onlay. For planar flaws in the balance of the dissimilar metal weld examination volume, the thickness ‘‘t’’ in IWB–3514 is the combined thickness of the inlay or onlay and the dissimilar metal weld. (8) Optimized weld overlay examination: Initial inservice examination of Inspection Item C–2 welds, shall be performed between the third refueling outage and no later than 10 years after application of the overlay. (9) Deferral: Note (11)(b)(1) in ASME BPV Code Case N–770–2 shall not be used to defer the initial inservice PO 00000 Frm 00046 Fmt 4701 Sfmt 9990 examination of optimized weld overlays (i.e., Inspection Item C–2 of ASME BPV Code Case N–770–2). (10) Examination technique: Note 14(b) of Table 1 and Note (b) of Figure 5(a) of ASME BPV Code Case N–770–2 may only be implemented if the requirements of Note 14(a) of Table 1 of ASME BPV Code Case N–770–2 cannot be met. (11) Cast stainless steel: Examination of ASME Code Class 1 piping and vessel nozzle butt welds involving cast stainless steel materials, shall be performed with Appendix VIII, Supplement 9 qualifications, or qualifications similar to Appendix VIII, Supplement 2 or 10 using cast stainless steel mockups no later than the next scheduled weld examination after January 1, 2020, in accordance with the requirements of paragraph-2500(a). (12) Stress improvement inspection coverage: Under Paragraph I.5.1, for cast stainless steel items, the required examination volume shall be examined by Appendix VIII procedures to the maximum extent practical including 100 percent of the susceptible material volume. (13) Encoded ultrasonic examination: Ultrasonic examinations performed in accordance with the requirements of Table 1 for Inspection Item A–1, A–2, B, E, F–2, J, and K shall be performed for essentially 100 percent of the inspection surface area using an encoded method. * * * * * Dated at Rockville, Maryland, this 21st day of August 2015. For the Nuclear Regulatory Commission. Michele G. Evans, Acting Director, Office of Nuclear Reactor Regulation. [FR Doc. 2015–23193 Filed 9–17–15; 8:45 am] BILLING CODE 7590–01–P E:\FR\FM\18SEP5.SGM 18SEP5

Agencies

NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 181 (Friday, September 18, 2015)]
[Proposed Rules]
[Pages 56819-56864]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-23193]



[[Page 56819]]

Vol. 80

Friday,

No. 181

September 18, 2015

Part V





Nuclear Regulatory Commission





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10 CFR Part 50





Incorporation by Reference of American Society of Mechanical Engineers 
Codes and Code Cases; Proposed Rule

Federal Register / Vol. 80 , No. 181 / Friday, September 18, 2015 / 
Proposed Rules

[[Page 56820]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[NRC-2011-0088]
RIN 3150-AI97


Incorporation by Reference of American Society of Mechanical 
Engineers Codes and Code Cases

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to 
amend its regulations to incorporate by reference seven recent editions 
and addenda to the American Society of Mechanical Engineers (ASME) 
codes for nuclear power plants and a standard for quality assurance. 
The NRC is also proposing to incorporate by reference four ASME code 
cases. This action is in accordance with the NRC's policy to 
periodically update the regulations to incorporate by reference new 
editions and addenda of the ASME codes and is intended to maintain the 
safety of nuclear power plants and to make NRC activities more 
effective and efficient.

DATES: Submit comments by December 2, 2015. Comments received after 
this date will be considered if it is practical to do so, but the NRC 
is able to ensure consideration only for comments received on or before 
this date.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2011-0088. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     Email comments to: Rulemaking.Comments@nrc.gov. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal 
workdays; telephone: 301-415-1677.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Daniel I. Doyle, Office of Nuclear 
Reactor Regulation, telephone: 301-415-3748, email: 
Daniel.Doyle@nrc.gov; or Keith Hoffman, Office of Nuclear Reactor 
Regulation, telephone: 301-415-1294, email: Keith.Hoffman@nrc.gov. Both 
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.

SUPPLEMENTARY INFORMATION: 

Executive Summary

A. Need for the Regulatory Action

    The NRC is proposing to amend its regulations to incorporate by 
reference seven recent editions and addenda to the ASME codes for 
nuclear power plants and an ASME standard for quality assurance. The 
NRC is also proposing to incorporate by reference four ASME code cases.
    This proposed rule is the latest in a series of rulemakings to 
amend the NRC's regulations to incorporate by reference revised and 
updated ASME codes for nuclear power plants. The ASME periodically 
revises and updates its codes for nuclear power plants by issuing new 
editions and addenda, and this rulemaking is in accordance with the 
NRC's policy to update the regulations to incorporate by reference 
those new editions and addenda. The incorporation by reference of the 
new editions and addenda will maintain the safety of nuclear power 
plants, make NRC activities more effective and efficient, and allow 
nuclear power plant licensees and applicants to take advantage of the 
latest ASME codes. The ASME is a voluntary consensus standards 
organization, and the ASME codes are voluntary consensus standards. The 
NRC's use of the ASME codes is consistent with applicable requirements 
of the National Technology Transfer and Advancement Act. Additional 
discussion of voluntary consensus standards and the NRC's compliance 
with the National Technology Transfer and Advancement Act (NTTAA) is 
set forth in Section VIII of this notice, ``Voluntary Consensus 
Standards.''

B. Major Provisions

    Major provisions of the proposed rule include:
     Incorporation by reference of ASME codes into NRC 
regulations and delineation of NRC requirements for the use of these 
codes (including conditions).
     Incorporation by reference of various versions of quality 
assurance standard NQA-1 into NRC regulations and approval for their 
use.
     Incorporation by reference and approval of four ASME Code 
Cases.

C. Costs and Benefits

    The NRC prepared a draft regulatory analysis to determine the 
expected costs and benefits of the proposed rule. The regulatory 
analysis identified costs and benefits in a quantitative fashion as 
well as in a qualitative fashion.
    The analysis concluded that the proposed rule would result in net 
quantitative costs to the industry and the NRC. The proposed rule, 
relative to the regulatory baseline, would result in a net cost for 
industry of between $5.1 million based on a 7 percent net present value 
and $4.3 million based on a 3 percent net present value. The estimated 
incremental industry cost per reactor unit ranges from $49,000 based on 
a 7 percent net present value to $41,000 based on a 3 percent net 
present value. The NRC benefits from the proposed rulemaking 
alternative because of the averted cost of not reviewing and approving 
Code alternative requests on a plant-specific basis under Sec.  
50.55a(z) of title 10 of the Code of Federal Regulations (10 CFR). The 
NRC net benefit ranges from $1.4 million based on a 7 percent net 
present value to $1.9 million based on a 3 percent net present value.
    Qualitative factors which were considered include regulatory 
stability and predictability, regulatory efficiency, and consistency 
with the NTTAA Act of 1995, as amended. Table 44 in the draft 
regulatory analysis includes a discussion of the costs and benefits 
that were considered qualitatively. If the results of the regulatory 
analysis were based solely on quantified costs and benefits, then the 
regulatory analysis would show that the rulemaking is not justified 
because the total quantified benefits of the proposed regulatory action 
do not equal or exceed the costs of the proposed action. However, if 
the qualitative benefits (including the safety benefit, cost savings, 
and other non-quantified benefits) are considered together with the 
quantified benefits, then the benefits outweigh the identified 
quantitative and qualitative impacts.
    With respect to regulatory stability and predictability, the NRC 
has had a decades-long practice of approving and/

[[Page 56821]]

or mandating the use of certain parts of editions and addenda of these 
ASME Codes in 10 CFR 50.55a through the rulemaking process of 
``incorporation by reference.'' Retaining the practice of approving 
and/or mandating the ASME Codes continues the regulatory stability and 
predictability provided by the current practice. Retaining the practice 
also assures consistency across the industry, and provides assurance to 
the industry and the public that the NRC will continue to support the 
use of the most updated and technically sound techniques developed by 
the ASME to provide adequate protection to the public. In this regard, 
these ASME Codes are voluntary consensus standards developed by 
participants with broad and varied interests and have already undergone 
extensive external review before being reviewed by the NRC. Finally, 
the NRC's use of the ASME Codes is consistent with the NTTAA, which 
directs Federal agencies to adopt voluntary consensus standards instead 
of developing ``government-unique'' (i.e., Federal agency-developed) 
standards, unless inconsistent with applicable law or otherwise 
impractical.
    For more information, please see the draft regulatory analysis 
(Accession No. ML14170B104 in the NRC's Agencywide Documents Access and 
Management System).

Table of Contents

I. Obtaining Information and Submitting Comments
    A. Obtaining Information
    B. Submitting Comments
II. Background
III. Discussion
    A. ASME BPV Code, Section III
    B. ASME BPV Code, Section XI
    C. ASME OM Code
    D. ASME Code Cases
IV. Section-by-Section Analysis
V. Generic Aging Lessons Learned Report
VI. Specific Request for Comments
VII. Plain Writing
VIII. Voluntary Consensus Standards
IX. Incorporation by Reference--Reasonable Availability to 
Interested Parties
X. Environmental Assessment and Final Finding of No Significant 
Environmental Impact
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis: Availability
XIII. Backfitting and Issue Finality
XIV. Regulatory Flexibility Certification
XV. Availability of Documents

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2011-0088 when contacting the NRC 
about the availability of information for this proposed rule. You may 
obtain information related to this proposed rule by any of the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2011-0088.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. For 
the convenience of the reader, instructions about obtaining materials 
referenced in this document are provided in the ``Availability of 
Documents'' section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2011-0088 in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Background

    The ASME develops and publishes the ASME Boiler and Pressure Vessel 
Code (BPV Code), which contains requirements for the design, 
construction, and inservice inspection (ISI) of nuclear power plant 
components; and the ASME OM Code,\1\ which contains requirements for 
inservice testing (IST) of nuclear power plant components. Until 2012, 
the ASME issued new editions of the ASME BPV Code every 3 years and 
addenda to the editions annually, except in years when a new edition 
was issued. Similarly, the ASME periodically published new editions and 
addenda of the ASME OM Code. Starting in 2012, the ASME decided to 
issue editions of its BPV and OM Codes (no addenda) every 2 years with 
the BPV Code to be issued on the odd years (e.g., 2013, 2015, etc.) and 
the OM Code to be issued on the even years (e.g., 2012, 2014, etc.). 
The new editions and addenda typically revise provisions of the Codes 
to broaden their applicability, add specific elements to current 
provisions, delete specific provisions, and/or clarify them to narrow 
the applicability of the provision. The revisions to the editions and 
addenda of the Codes do not significantly change Code philosophy or 
approach.
---------------------------------------------------------------------------

    \1\ The editions and addenda of the ASME Code for Operation and 
Maintenance of Nuclear Power Plants have had different titles from 
2005 to 2012 and are referred to collectively in this rule as the 
``OM Code.''
---------------------------------------------------------------------------

    It has been the NRC's practice to establish requirements for the 
design, construction, operation, ISI (examination), and IST of nuclear 
power plants by approving the use of editions and addenda of the ASME 
BPV and OM Codes (ASME Codes) in Sec.  50.55a. The NRC approves and/or 
mandates the use of certain parts of editions and addenda of these ASME 
Codes in Sec.  50.55a through the rulemaking process of ``incorporation 
by reference.'' Upon incorporation by reference of the ASME Codes into 
Sec.  50.55a, the provisions of the ASME Codes are legally-binding NRC 
requirements as delineated in Sec.  50.55a, and subject to the 
conditions on certain specific ASME Codes' provisions that are set 
forth in Sec.  50.55a. The editions and addenda of the ASME BPV and OM 
Codes were last incorporated by reference into the regulations in a 
final rule dated June 21, 2011 (76 FR 36232), subject to NRC 
conditions.
    The ASME Codes are consensus standards developed by participants 
with broad and varied interests (including the NRC and licensees of 
nuclear power plants). The ASME's adoption of new editions of, and 
addenda to, the ASME Codes does not mean that there is unanimity on 
every provision in the ASME Codes. There may be disagreement among the 
technical experts, including NRC representatives on the ASME Code 
committees and subcommittees, regarding the acceptability or 
desirability of a particular Code

[[Page 56822]]

provision included in an ASME-approved code edition or addenda. If the 
NRC believes that there is a significant technical or regulatory 
concern with a provision in an ASME-approved Code edition or addenda 
being considered for incorporation by reference, then the NRC 
conditions the use of that provision when it incorporates by reference 
that ASME Code edition or addenda. In some cases, the condition 
increases the level of safety afforded by the ASME code provision, or 
addresses a regulatory issue not considered by the ASME. In other 
instances, where research data or experience has shown that certain 
Code provisions are unnecessarily conservative, the condition may 
provide that the Code provision need not be complied with in some or 
all respects. The NRC's conditions are included in Sec.  50.55a, 
typically in paragraph (b) of that regulation. In a Staff Requirements 
Memorandum (SRM) dated September 10, 1999, the Commission indicated 
that NRC rulemakings adopting (incorporating by reference) a voluntary 
consensus standard must identify and justify each part of the standard 
that is not adopted. For this rulemaking, the provisions of the 2009 
Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section III, 
Division 1; and the 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 
Edition of Section XI, Division 1, of the ASME BPV Code; and the 2009 
Edition, 2011 Addenda, and 2012 Edition of the ASME OM Code that the 
NRC is not adopting, or partially adopting, are identified in the 
Discussion, Regulatory Analysis, and Backfitting and Issue Finality 
sections of this notice. The provisions of those specific editions and 
addenda and Code Cases that are the subject of this rulemaking that the 
NRC finds to be conditionally acceptable, together with the applicable 
conditions, are also identified in the Discussion, Regulatory Analysis, 
and Backfitting and Issue Finality sections of this notice.
    The ASME Codes are voluntary consensus standards, and the NRC's 
incorporation by reference of these Codes is consistent with applicable 
requirements of the NTTAA. Additional discussion on NRC's compliance 
with the NTTAA is set forth in Section VIII of this notice, ``Voluntary 
Consensus Standards.''
    This proposed rule contains changes from a November 5, 2014, NRC 
final rule amending Sec.  50.55a to, among other things, re-designate 
paragraphs within Sec.  50.55a (79 FR 65776). The re-designation of 
paragraphs was needed to address the Office of the Federal Register's 
requirements in 10 CFR part 51 applicable to incorporation by 
reference. For additional information on the November 2014 final rule, 
please consult the statement of considerations (preamble) for that 
final rule.

III. Discussion

    The NRC regulations incorporate by reference ASME codes for nuclear 
power plants. The ASME periodically revises and updates its codes for 
nuclear power plants. This proposed rule is the latest in a series of 
rulemakings to amend the NRC's regulations to incorporate by reference 
revised and updated ASME codes for nuclear power plants. This 
rulemaking is intended to maintain the safety of nuclear power plants 
and make NRC activities more effective and efficient.
    The NRC follows a three-step process to determine acceptability of 
new provisions in new editions and addenda to the Codes and the need 
for conditions on the uses of these Codes. This process was employed in 
the review of the Codes that are the subjects of this rule. First, the 
NRC staff actively participates with other ASME committee members with 
full involvement in discussions and technical debates in the 
development of new and revised Codes. This includes a technical 
justification of each new or revised Code. Second, the NRC committee 
representatives discuss the Codes and technical justifications with 
other cognizant NRC staff to ensure an adequate technical review. 
Third, the NRC position on each Code is reviewed and approved by NRC 
management as part of the rule amending Sec.  50.55a to incorporate by 
reference new editions and addenda of the ASME Codes and conditions on 
their use. This regulatory process, when considered together with the 
ASME's own process for developing and approving the ASME Codes, 
provides reasonable assurance that the NRC approves for use only those 
new and revised Code edition and addenda, with conditions as necessary, 
that provide reasonable assurance of adequate protection to public 
health and safety, and that do not have significant adverse impacts on 
the environment.
    The NRC reviewed changes to the Codes in the editions and addenda 
of the Codes identified in this rulemaking. The NRC concluded, in 
accordance with the process for review of changes to the Codes, that 
each of the editions and addenda of the Codes, and the 2008 Edition and 
the 2009-1a Addenda of NQA-1, are technically adequate, consistent with 
current NRC regulations, and approved for use with the specified 
conditions.
    The NRC proposes to amend its regulations to incorporate by 
reference:
     The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 
Edition to the ASME BPV Code, Section III, Division 1 and Section XI, 
Division 1, with conditions on their use.
     The 2009 Edition, the 2011 Addenda, and the 2012 Edition 
to Division 1 of the ASME OM Code, with conditions on their use.
     ASME Standard NQA-1, ``Quality Assurance Requirements for 
Nuclear Facility Applications,'' including several editions and addenda 
to NQA-1 from previous years with slightly varying titles as identified 
in proposed rule language Sec.  50.55a(a)(1)(v). More specifically, the 
NRC proposes to incorporate by reference the 1983 Edition through the 
1994 Edition, the 2008 Edition, and the 2009-1a Addenda to the 2008 
Edition of ASME NQA-1, with conditions on their use.
     ASME BPV Code Case N-729-4, ``Alternative Examination 
Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having 
Pressure-Retaining Partial-Penetration Welds Section XI, Division 1,'' 
ASME approval date: June 22, 2012, with conditions on its use.
     ASME BPV Code Case N-770-2, ``Alternative Examination 
Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel 
Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler 
Material With or Without Application of Listed Mitigation Activities, 
Section XI, Division 1,'' ASME approval date: June 9, 2011, with 
conditions on its use.
     ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast 
Austenitic Piping Welds From the Outside Surface Section XI, Division 
1,'' ASME approval date: October 16, 2012.
     ASME OM Code Case OMN-20, ``Inservice Test Frequency.''
    The current regulations in Sec.  50.55a(a)(1)(ii) incorporate by 
reference ASME BPV Code, Section XI, 1970 Edition through the 1976 
Winter Addenda; and the 1977 Edition (Division 1) through the 2008 
Addenda (Division 1), subject to the conditions identified in current 
Sec.  50.55a(b)(2)(i) through (b)(2)(xxix). The proposed amendment 
would revise Sec.  50.55a(a)(1)(ii) to incorporate by reference the 
2009 Addenda (Division 1) through the 2013 Edition (Division 1) of the 
ASME BPV Code, Section XI. It would also clarify the wording and add, 
remove, or revise some of the conditions as explained in this notice.
    The NRC proposes to revise Sec.  50.55a(a)(1)(iv) to incorporate by 
reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division 
1 of the ASME OM Code. Based on this revision, the NRC regulations 
would

[[Page 56823]]

incorporate by reference in Sec.  50.55a the 1995 Edition through the 
2012 Edition of the ASME OM Code.
    Each of the proposed NRC conditions and the reasons for each 
proposed condition are discussed below. The discussions are organized 
under the applicable ASME Code and Section. Please note that there is 
not a separate heading for ASME quality assurance standard NQA-1 
because there are three separate discussions of NQA-1--one under the 
heading for ASME BPV Code, Section III, one under the heading for ASME 
BPV Code, Section XI, and one under the heading for ASME OM Code--
because there are three proposed conditions related to NQA-1, one in 
each of those areas (paragraph (b)(1)(iv) for Section III, paragraph 
(b)(2)(x) for Section XI, and paragraph (b)(3)(i) for the OM Code).

A. ASME BPV Code, Section III

10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section 
III
    The NRC proposes to clarify that Section III Nonmandatory 
Appendices are not incorporated by reference. This language was 
originally added in a final rule published on June 21, 2011 (76 FR 
36232); however, it was omitted from the final rule published on 
November 5, 2014 (79 FR 65776). The NRC is correcting the omission by 
inserting ``(excluding Non-mandatory Appendices)'' in 10 CFR 
50.55a(a)(1)(i).
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
    The NRC proposes to identify prohibited subparagraphs and footnotes 
for each BPV Code edition and addenda in tabular form as opposed to the 
textual listing of the current regulation. No substantive change to the 
requirements is intended by this revision. The NRC believes that 
presenting the information in tabular form will increase the clarity 
and understandability of the regulation.
    Currently, Sec.  50.55a(b)(1)(ii) includes a condition prohibiting 
the use of Footnote 11 from the 1989 Addenda through the 2003 Addenda 
or Footnote 13 from the 2004 Edition through the 2008 Addenda to 
Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg sizes less 
than 1.09 tn when using the ASME BPV Code, Section III, 
Division 1. These Code provisions provide stress indices for welded 
joints used in the design of Class 2 and Class 3 piping. The use of 
these indices is prohibited for welds with leg sizes less than 1.09 
tn, where tn is the nominal pipe thickness. This 
is due to the fact that the current provisions would result in a weld 
that would be weaker than the pipe to which it is adjoined under these 
dimensions. The weld stress provisions in the version of the footnotes 
contained in the 1989 Addenda have been relocated to different 
subparagraphs in subsequent BPV Code editions and addenda. Therefore, 
the current Code's reference in Footnote 11 to Figures NC-3673.2(b)-1 
and ND-3673.2(b)-1 is not correct for BPV Code editions and addenda 
after the 1989 Addenda, in applying the condition. The proposed rule 
would correct this issue by clearly identifying the prohibited code 
provisions in the editions and addenda in a tabular format.
    As an editorial matter, this proposed rule identifies the 
prohibited BPV Code provisions as ``notes,'' which is the term used by 
the ASME, rather than ``footnotes.'' The NRC proposes to use the 
terminology used by the ASME for clarity.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
    The NRC proposes to approve for use the version of NQA-1 referenced 
in the 2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV 
Code, Section III, Subsection NCA, Article 7000, which this rule is 
also incorporating by reference. This will allow applicants and 
licensees to use the 2008 Edition and the 2009-1a Addenda of NQA-1 when 
using the 2010 and later editions and addenda of Section III.
    In the 2010 Edition of ASME BPV Code, Section III, Subsection NCA, 
Article NCA-4000, ``Quality Assurance,'' was updated to require N-Type 
Certificate Holders to comply with the requirements of Part 1 of the 
2008 Edition and the 2009-1a Addenda of ASME Standard NQA-1, ``Quality 
Assurance Requirements for Nuclear Facility Applications,'' as modified 
and supplemented in NCA-4120(b) and NCA-4134. In addition, NCA-4110(b) 
was revised to remove the reference to a specific edition and addenda 
of ASME NQA-1, and Table NCA-7100-2, ``Standards and Specifications 
Referenced in Division 1,'' was updated to require the 2008 Edition and 
2009-1a Addenda of NQA-1 when using the 2010 Edition of Section III.
    The NRC reviewed the 2008 Edition and the 2009-1a Addenda of NQA-1 
and compared it to previously approved versions of NQA-1 and found that 
there were no significant differences. In addition, the NRC reviewed 
the changes to Subsection NCA that reference the 2008 Edition and 2009-
1a Addenda of NQA-1, compared them to previously approved versions of 
Subsection NCA, and found that there were no significant differences. 
Therefore, the NRC has concluded that these Editions and Addenda of 
NQA-1 are acceptable for use.
    The NRC proposes to revise Sec.  50.55a(b)(1)(iv) to clarify that 
an applicant's or licensee's commitments, addressing those areas where 
NQA-1 either does not address a requirement in appendix B to 10 CFR 
part 50, ``Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants,'' or is less stringent than the comparable 
appendix B requirement, governs the applicant's or licensee's Section 
III activities. The proposed clarification is consistent with Sec.  
50.55a(b)(2)(x) and Sec.  50.55a(b)(3)(i). NQA-1 provides the ASME's 
method for establishing and implementing a quality assurance (QA) 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants. However, NQA-1, as modified and supplemented 
in NCA-4120(b) and NCA-4134, does not address some of the requirements 
of appendix B to 10 CFR part 50. In some cases, the provisions of NQA-1 
are less stringent than the comparable appendix B requirement. Thus, in 
order to meet the requirements of appendix B, an applicant's or 
licensee's QA program description must contain commitments addressing 
those provisions of appendix B which are not covered by NQA-1, as well 
as provisions that supplement or replace the NQA-1 provisions where the 
appendix B requirement is more stringent.
    Finally, the NRC is considering removing the reference in Sec.  
50.55a(b)(1)(iv) to versions of NQA-1 older than the 1994 Edition. The 
NRC requests public comment on whether any applicant or licensee is 
committed to, and is using, a version of NQA-1 older than the 1994 
Edition, and if so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification 
and Demonstration of Function of Incompressible-Fluid Pressure-Relief 
Valves
    The NRC proposes to revise Sec.  50.55a(b)(1)(vii) so that the 
existing condition prohibiting the use of paragraph NB-7742(a)(2) of 
the 2006 Addenda through the 2007 Edition up to and including the 2008 
Addenda is extended to include the editions and addenda up to the 2013 
Edition which are the subject of this rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME 
Certification Marks
    The NRC is proposing to add new paragraph, Sec.  
50.55a(b)(1)(viii), to allow

[[Page 56824]]

licensees to use either the ASME BPV Code Symbol Stamps of editions and 
addenda earlier than the 2011 Addenda to the 2010 Edition of the ASME 
BPV Code or the ASME Certification Marks with the appropriate 
certification designators and class designators as specified in the 
2013 Edition through the latest edition and addenda incorporated by 
reference in 10 CFR 50.55a.
    The ASME BPV Code requires, in certain instances, that components 
be stamped. The stamp signifies that the component has been designed, 
fabricated, examined and tested, as specified in the ASME BPV Code. The 
stamp also signifies that the required ASME BPV Code data report forms 
have been completed, and the authorized inspector has inspected the 
item and authorized the application of the ASME BPV Code Symbol Stamp.
    The ASME has instituted changes in the BPV Code to consolidate the 
different ASME BPV Code Symbol Stamps into a common ASME Certification 
Mark. This action was implemented in the 2011 Addenda to the 2010 
Edition of the ASME BPV Code. As of the end of 2012, ASME no longer 
utilizes the ASME BPV Code Symbol Stamp. Licensees, however, may not 
have updated to the Edition or Addenda that identifies the use of the 
ASME Certification Mark. Nevertheless, licensees are legally required 
to implement the ASME BPV Code Edition and Addenda identified as their 
current code of record. As ASME components are procured, these 
components may be received with the ASME Certification Mark, while the 
licensee's current code of record may require the component to have the 
ASME BPV Code Symbol Stamp. Installation of a component under such 
circumstances would not be in compliance with the regulations that the 
licensees are required to meet.
    Both the ASME Certification Mark and the ASME BPV Code Symbol Stamp 
are official ASME methods of certifying compliance with the Code. 
Although these ASME Certification Marks differ slightly in appearance, 
they serve the same purpose of certifying code compliance by the ASME 
Certificate Holder and continue to provide for the same level of 
quality assurance for the application of the ASME Certification Mark as 
was required for the application of the ASME BPV Code Symbol Stamp. The 
new ASME Certification Mark represents a small, non-safety significant 
modification of ASME's trademark. As such, it does not change the 
technical requirements of the Code. ASME has confirmed that the 
Certification Mark with designator is equivalent to the corresponding 
BPV Code Symbol Stamp. Based on statements by ASME in a letter dated 
August 17, 2012, the NRC has concluded that the ASME BPV Code Symbol 
Stamps and ASME Certification Mark with code-specific designators are 
equivalent with respect to their certification of compliance with the 
BPV Code. The NRC discussed this issue in Regulatory Issue Summary 
2013-07, ``NRC Staff Position on the Use of American Society of 
Mechanical Engineers Certification Mark,'' dated May 28, 2013.

B. ASME BPV Code, Section XI

10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section 
XI
    The NRC proposes to revise Sec.  50.55a(a)(1)(ii) to clarify that 
Section XI Non-mandatory Appendix U of the 2013 Edition of ASME BPV 
Code Section XI is not incorporated by reference and therefore not 
approved for use. The NRC is developing an integrated approach to the 
issue of operational leakage. The NRC has not completed its 
determination of how Appendix U fits into this integrated approach to 
address the operational leakage issue at nuclear power plants. The 
operational leakage issue has many factors that need to be considered 
such as acceptance criteria, corrective actions, application of repair/
replacement requirements, component operability determination, concerns 
related to continued operation, maximum acceptable leakage rates, flaw 
growth rates, flaw measurement techniques, schedules for eliminating 
leakage, and when or if the leakage requires authorization by the NRC. 
The NRC plans to complete the development of the regulatory approach to 
operational leakage and issue it in a future rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and 
Addenda of Subsection IWE and Subsection IWL
    The NRC proposes to revise Sec.  50.55a(b)(2)(vi) to explicitly 
state that the provision requiring the use of either the 1992 Edition 
with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of 
Subsection IWE and Subsection IWL when implementing the initial 120-
month containment inservice inspection program applies only to those 
licensees that were required by previous versions \2\ of Sec.  50.55a 
to develop and implement a containment inservice inspection program in 
accordance with Subsection IWE and Subsection IWL, and complete an 
expedited examination of containment during the 5-year period from 
September 9, 1996, to September 9, 2001.
---------------------------------------------------------------------------

    \2\ See the supplementary information and rule language for 
Sec.  50.55a(b)(2)(vi), Sec.  50.55a(g)(4), and Sec.  
50.55a(g)(6)(ii)(B) in Federal Register notices published on August 
8, 1996 (61 FR 41303), and September 22, 1999 (64 FR 51370).
---------------------------------------------------------------------------

    The expedited examination involved the completion of the first set 
of examinations of the first or initial 120-month containment 
inspection interval. It is noted that all the operating reactors in the 
above stated class would have gone past their initial 120-month 
inspection interval by 2011. The proposed change removes the 
possibility of misinterpretation of the provision as requiring plants 
that do not fall in the above class, such as reactors licensed after 
September 9, 2001, to use the 1992 Edition with 1992 Addenda or the 
1995 Edition with 1996 Addenda of Subsection IWE and Subsection IWL, 
Section XI for implementing the initial 120-month inspection interval 
of the containment inservice inspection program. Applicants and 
licensees that do not fall in the above class must use Code editions 
and addenda in accordance with Sec.  50.55a(g)(4)(i) and (g)(4)(ii), 
respectively, for the initial and successive 120-month containment 
inservice inspection intervals.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment 
Examinations
    The NRC proposes to revise Sec.  50.55a(b)(2)(viii) by removing the 
condition for using the 2007 Edition with 2009 Addenda through the 2013 
Edition of Subsection IWL requiring compliance with Sec.  
50.55a(b)(2)(viii)(E) and adding a requirement to comply with Sec.  
50.55a(b)(2)(viii)(H) and (I).
    Section 50.55a(b)(2)(viii)(E) is one of several conditions that 
apply to the inservice examination of concrete containments using 
Subsection IWL of various editions and addenda of the ASME BPV Code, 
Section XI, incorporated by reference in Sec.  50.55a(a)(1)(ii). The 
NRC proposes to remove the condition in Sec.  50.55a(b)(2)(viii)(E) 
when applying the 2007 Edition with 2009 Addenda through the 2013 
Edition of Subsection IWL because its intent has been incorporated into 
the Code in the new provision IWL-2512, ``Inaccessible Areas.'' The 
reasons for requiring compliance with Sec.  50.55a(b)(2)(viii)(H) and 
(I) are set forth in the next two sections.

[[Page 56825]]

10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth 
Provision
    The NRC proposes to add a new paragraph, Sec.  
50.55a(b)(2)(viii)(H), to specify the information that must be provided 
in the ISI Summary Report required by IWA-6000, when inaccessible 
concrete surfaces are evaluated under the new code provision IWL-2512. 
This new condition would replace the existing condition in Sec.  
50.55a(b)(2)(viii)(E) when using the 2007 Edition with the 2009 Addenda 
through the 2013 Edition of Subsection IWL.
    The existing condition in Sec.  50.55a(b)(2)(viii)(E) of the 
current rule requires that, for Class CC applications, the licensee 
shall evaluate the acceptability of inaccessible areas when conditions 
exist in accessible areas that could indicate the presence of or result 
in degradation to such inaccessible areas, and provide the evaluation 
information required by Sec. Sec.  50.55a(b)(2)(viii)(E)(1), (E)(2), 
and (E)(3) in the IWA-6000 ISI Summary Report.
    In the 2009 Addenda Subsection IWL, the ASME revised existing 
provisions IWL-1220 and IWL-2510 and added new provision IWL-2512 
intended to incorporate the condition in Sec.  50.55a(b)(2)(viii)(E) 
into Subsection IWL. The IWL-2510, ``Surface Examination,'' was 
restructured into new paragraphs IWL-2511, ``Accessible Areas,'' with 
almost the same provisions as the previous IWL-2510 and IWL-2512, 
``Inaccessible Areas,'' to be specific to examinations required for 
accessible areas, and differentiate between those and the new 
requirements for inaccessible areas. The inaccessible areas addressed 
by the new IWL-2512 are: (1) Concrete surfaces obstructed by adjacent 
structures, parts or appurtenances (e.g., generally above-grade 
inaccessible areas) and (2) concrete surfaces made inaccessible by 
foundation material or backfill (e.g., below-grade inaccessible areas).
    The revised IWL-2511(a) has a new requirement that states that, 
``If the Responsible Engineer determines that observed suspect 
conditions indicate the presence of, or could result in, degradation of 
inaccessible areas, the requirements of IWL-2512(a) shall be met.'' The 
new IWL-2512(a) requires the ``Responsible Engineer'' to evaluate 
suspect conditions and specify the type and extent of examinations, if 
any, required to be performed on inaccessible surface areas described 
in the previous paragraph. The acceptability of the evaluated 
inaccessible area would be determined either based on the evaluation or 
based on the additional examinations, if determined to be required. The 
new IWL-2512(b) further requires a periodic technical evaluation of 
below-grade inaccessible areas of concrete to be performed to determine 
and manage its susceptibility to degradation regardless of whether 
suspect conditions exist in accessible areas that would warrant an 
evaluation of inaccessible areas based on the condition in Sec.  
50.55a(b)(2)(viii)(E). Therefore, the revised IWL-2511(a) and new IWL-
2512 code provisions address the evaluation and acceptability of 
inaccessible areas consistent with the existing condition in Sec.  
50.55a(b)(2)(viii)(E), with one exception. The exception is that the 
new IWL-2512 provision does not explicitly require the information 
specified in Sec. Sec.  50.55a(b)(2)(viii)(E)(1), (E)(2), and (E)(3) of 
the existing condition to be provided in the IWA-6000 ISI Summary 
Report.
    For these reasons, the NRC proposes to identify the information 
that must be provided in the ISI Summary Report required by IWA-6000 
when inaccessible concrete surfaces are evaluated under the new code 
provision IWL-2512. This new condition would replace the existing 
condition in Sec.  50.55a(b)(2)(viii)(E) when using the 2007 Edition 
with the 2009 Addenda through the 2013 Edition of Subsection IWL. The 
information requested by the new condition must be provided when 
inaccessible concrete areas are evaluated per IWL-2512(a) for 
degradation based on suspect conditions found in accessible areas, as 
well as when periodic technical evaluations of inaccessible below-grade 
concrete areas required by IWL-2512(b) are performed.
10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth 
Provision
    The NRC proposes to add Sec.  50.55a(b)(2)(viii)(I) to place a 
condition on the periodic technical evaluation requirements in the new 
IWL-2512(b), for consistency with NUREG-1801, Revision 2, ``Generic 
Aging Lessons Learned (GALL) Report,'' with regard to aging management 
of below-grade containment concrete surfaces. The new IWL-2512(b) 
provision is applicable to inaccessible below-grade concrete surfaces 
exposed to foundation soil, backfill, or groundwater. This condition 
would apply only during the period of extended operation of a renewed 
license under 10 CFR part 54, when using IWL-2512(b) of the 2007 
Edition with 2009 Addenda through the 2013 Edition of Subsection IWL.
    In the 2009 Addenda of Subsection IWL, the ASME added new code 
provisions, IWL-2512(b) and (c) as well as a new line item L1.13 in 
Table IWL-2500-1, intended to specifically address aging management 
concerns with potentially unidentified degradation of inaccessible 
below-grade containment concrete areas and to be responsive to actions 
outlined in the GALL Report related to aging management of inaccessible 
below-grade concrete surfaces. It is noted that these new code 
provisions are an enhancement to the requirement of the existing 
condition in Sec.  50.55a(b)(2)(viii)(E) to specifically address aging 
management of inaccessible below-grade containment concrete areas and 
is generally acceptable to the NRC.
    The new IWL-2512(b) provides requirements for systematically 
performing a periodic technical evaluation of concrete surfaces exposed 
to foundation soil, backfill, or groundwater to determine 
susceptibility of the concrete to deterioration that could affect its 
ability to perform its intended design function under conditions 
anticipated through the service life of the structure. It requires the 
technical evaluation to be performed and documented at periodic 
intervals not to exceed 10 years regardless of whether conditions exist 
in accessible areas that would warrant an evaluation of inaccessible 
areas by the existing condition in Sec.  50.55a(b)(2)(viii)(E), which 
the NRC finds reasonable for the initial 40-year operating license 
period. The new IWL-2512(b) further provides the specific elements, 
including aging mechanisms considered, that the technical evaluation 
should include, as well as the definition of an aggressive below-grade 
environment. The new IWL-2512(c) requires that the evaluation results 
of IWL-2512(b) be used to define and document the condition monitoring 
program, if determined to be required, including required examinations 
and frequencies, to be implemented for the management of degradation 
and aging effects of the below-grade concrete surface areas. If it is 
determined that additional examinations are required, these 
examinations of inaccessible below-grade areas will be implemented in 
accordance with new line item L1.13 in Table IWL-2500-1 under 
Examination Category L-A, Concrete, with acceptance criteria based on 
IWL-3210. It should be noted that a technical evaluation approach, such 
as in IWL-2512(b), could be used, and is generally used, to determine 
acceptability of a

[[Page 56826]]

below-grade inaccessible area to satisfy the condition in Sec.  
50.55a(b)(2)(viii)(E).
    The technical evaluation requirements in IWL-2512(b) help to 
determine the susceptibility to degradation and manage aging effects of 
inaccessible below-grade concrete surfaces, before the loss of intended 
function. The requirements are based on, and are generally consistent 
with, the guidance in the GALL Report,'' with the following two 
exceptions. The first exception is that IWL-2512(b) requires the 
technical evaluation to determine the susceptibility of the concrete to 
degradation and the ability to perform the intended design function 
through its service life at periodic intervals not to exceed 10 years. 
The aging management programs (AMPs) for safety-related structures 
(e.g., Structures Monitoring) in the GALL Report require such 
evaluation to be performed at intervals not to exceed 5 years, which is 
also consistent with applicant commitments during review of license 
renewal applications. The second exception is that IWL-2512(b) requires 
that examination of representative samples of below-grade concrete be 
performed if excavated for any reason when an aggressive below-grade 
environment is present. However, the AMPs (X1.S6 Structures Monitoring 
and X1.S7 Water Control Structures) in the GALL Report require the same 
examination even for a non-aggressive below-grade environment.
    Based on these reasons, the NRC proposes to add a new Sec.  
50.55a(b)(2)(viii)(I) to place a condition on the periodic technical 
evaluation requirements in IWL-2512(b) for consistency with the GALL 
Report, with regard to aging management of inaccessible below-grade 
concrete components of the containment. The new IWL-2512(b) is 
applicable to inaccessible below-grade concrete surfaces of the 
containment cylindrical wall and basemat foundations, which are exposed 
to foundation soil, backfill, or groundwater. The new condition 
requires that, during the period of extended operation of a renewed 
license, the technical evaluation under IWL-2512(b) of inaccessible 
below-grade concrete surfaces exposed to foundation soil, backfill, or 
groundwater be performed at periodic intervals not to exceed 5 years. 
Also, the condition requires the examination of representative samples 
of the exposed portions of the below-grade concrete be performed when 
excavated for any reason. Since the GALL Report is the technical basis 
document for license renewal, this new condition applies only during 
the period of extended operation of a renewed license under 10 CFR part 
54, when using IWL-2512(b) of the 2007 Edition with 2009 Addenda 
through the 2013 Edition of Subsection IWL, Section XI.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment 
Examinations
    The NRC proposes to continue to apply the existing conditions in 
Sec. Sec.  50.55a(b)(2)(ix)(A)(2), (b)(2)(ix)(B), and (b)(2)(ix)(J) 
governing examinations of metal containments and the liners of concrete 
containments under Subsection IWE to the 2007 Edition with 2009 Addenda 
through the 2013 Edition (the code editions and addenda which are the 
subject of this rulemaking). The NRC reviewed the code changes in 
Subsection IWE of the 2009 Addenda through the 2013 Edition of ASME BPV 
Code, Section XI, and notes that all of the changes were editorial or 
administrative with the intent to improve the clarity of the existing 
requirements or correct errors by errata. There were no changes to 
Subsection IWE in the code editions and addenda that are the subject of 
this rulemaking that the NRC believes would require new regulatory 
conditions to ensure safety, nor do the changes to Subsection IWE 
address the NRC's reasons for adopting the conditions on the use of 
Subsection IWE. Although this continuation of the applicability of the 
three conditions does not require a rule change, the NRC is discussing 
this for the benefit of stakeholder understanding of the effect of the 
proposed rule.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
    The NRC proposes to approve for use the version of NQA-1 referenced 
in the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition 
of the ASME BPV Code, Section XI, Table IWA 1600-1, ``Referenced 
Standards and Specifications,'' which this rule is also incorporating 
by reference. This will allow licensees to use the 1994 or the 2008 
Edition and the 2009-1a Addenda of NQA-1 when using the 2009 Addenda 
and later editions and addenda of Section XI.
    In the 2013 Edition of ASME BPV Code, Section XI, Table IWA 1600-1 
was updated to allow licensees to use the 1994 or the 2008 Edition with 
the 2009-1a Addenda of NQA-1 when using the 2013 Edition of Section XI. 
In the 2010 Edition of ASME BPV Code, Section XI, IWA-1400, ``Owner's 
Responsibilities,'' subparagraph (n)(2) was updated to reference the 
NQA-1 Part I, Basic Requirements and Supplementary Requirements for 
Nuclear Facilities. In the 2009 Addenda of the 2007 Edition of ASME BPV 
Code, Section XI, Table IWA-1600-1, ``Referenced Standards and 
Specifications,'' was updated to allow licensees to use the 1994 
Edition of NQA-1. The NRC reviewed the 2008 Edition and the 2009-1a 
Addenda of NQA-1 and compared it to previously approved versions of 
NQA-1 and found that there were no significant differences. Therefore, 
the NRC has concluded that these Editions and Addenda of NQA-1 are 
acceptable for use.
    The NRC proposes to amend Sec.  50.55a(b)(2)(x) to clarify that a 
licensee's commitments addressing those areas where NQA-1 either does 
not address an appendix B requirement or is less stringent than the 
comparable appendix B requirement governs the licensee's Section XI 
activities. The proposed clarification is consistent with Sec. Sec.  
50.55a(b)(1)(iv) and (b)(3)(i). The ASME's method for establishing and 
implementing a QA program for the design and construction of nuclear 
power plants and fuel reprocessing plants is described in NQA-1. 
However, NQA-1 does not address some of the requirements of appendix B 
to 10 CFR part 50. In some cases, the provisions of NQA-1 are less 
stringent than the comparable appendix B requirement. Thus, in order to 
meet the requirements of appendix B, a licensee's QA program 
description must contain commitments addressing those provisions of 
appendix B which are not covered by NQA-1, as well as provisions that 
supplement or replace the NQA-1 provisions where the appendix B 
requirement is more stringent.
    Finally, the NRC is considering removing the reference in Sec.  
50.55a(b)(2)(x) to versions of NQA-1 older than the 1994 Edition. The 
NRC requests public comment on whether any licensee is committed to, 
and is using, a version of NQA-1 older than the 1994 Edition, and if 
so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth 
Provision
    The NRC proposes to add a new paragraph, Sec.  
50.55a(b)(2)(xviii)(D), to prohibit applicants and licensees from using 
the ultrasonic examination nondestructive examination (NDE) personnel 
certification requirements in Section XI, Appendix VII and subarticle 
VIII-2200 of the 2011 Addenda and 2013 Edition of the ASME BPV Code. 
Section 50.55a(b)(2)(xviii) currently includes conditions on the 
certification

[[Page 56827]]

of NDE personnel. In addition, the new paragraph would require 
applicants and licensees to use the 2010 Edition, Table VII-4110-1 
training hour requirements for Levels I, II, and III ultrasonic 
examination personnel, and the 2010 Edition, subarticle VIII-2200 of 
Appendix VIII prerequisites for personnel requirements. In the 2011 
Addenda and 2013 Edition, the ASME BPV Code added an accelerated 
Appendix VII training process for certification of ultrasonic 
examination personnel based on training and prior experience, and 
separated the Appendix VII training requirements from the Appendix VIII 
qualification requirements. These new ASME BPV Code provisions would 
provide personnel in training with less experience and exposure to 
representative flaws in representative materials and configurations 
common to operating nuclear power plants, and they would permit 
personnel with prior non-nuclear ultrasonic examination experience to 
qualify for examinations in nuclear power plants without exposure to 
the variety of defects, examination conditions, components, and 
regulations common to operating nuclear power plants.
    The impact of reduced training and nuclear power plant 
familiarization is unknown. The ASME BPV Code supplants training hours 
and field experience without a technical basis, minimum defined 
training criteria, process details, or standardization. For these 
reasons, the NRC is proposing to prohibit the use of Appendix VII and 
VIII-2200 in the 2011 Addenda and 2013 Edition, and instead require 
applicants and licensees using the 2011 Addenda and 2013 Edition to use 
Table VII-4110-1 in the 2010 Edition, and VIII-2200, Appendix VIII 
prerequisites for ultrasonic examination personnel requirements in the 
2010 Edition.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements: 
First Provision
    The NRC proposes to revise Sec.  50.55a(b)(2)(xxi)(A) to modify the 
standard for visual magnification resolution sensitivity and contrast 
for visual examinations performed on Examination Category B-D 
components instead of ultrasonic examinations, making the rule conform 
with ASME BPV Code, Section XI requirements for VT-1 examinations. The 
character recognition rules are used in ASME BPV Code, Section XI, 
Table IWA-2211-1 for VT-1 tests, and are the standard tests used for 
resolution and contrast checks of VT-1 equipment. This revision 
essentially removes a requirement that was in addition to ASME BPV Code 
that required 1-mil wires to be used in licensees' Sensitivity, 
Resolution and Contrast Standard targets. In 2004, the NRC published 
NUREG/CR-6860, ``An Assessment of Visual Testing,'' showing that a 
linear target, such as a wire, is not an effective method for testing 
the resolution of a video camera system. In addition, BWRVIP-03 was 
changed to eliminate a \1/2\ mil wire from the Sensitivity Resolution 
and Contrast Standards due to similar concerns.
    Simple line detection can be a poor performance standard, allowing 
detection of a highly blurred image. This does not emulate sharpness 
quality recognition for evaluation of weld discontinuities. The 750 
[mu]m (30 mil) and the even smaller 25 [mu]m (1 mil) widths should not 
be used as performance standards because they do not determine image 
sharpness. This technique only measures the ``visible minimum'' for 
long linear indications, and does not measure a system's resolution or 
recognition limits. If the wire, or printed line, has a strong enough 
contrast against the background, then a linear feature well below the 
resolution of a system can be detected.
10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator 
Preservice Examinations
    The NRC proposes to add Sec.  50.55a(b)(2)(xxx) to require a full 
length examination of 100 percent of the tubing in each newly installed 
steam generator prior to plant startup. This requirement would be 
instead of the unapproved provisions in IWB-2200(c) pertaining to steam 
generator tube preservice inspections (PSI).
    Steam generator tubes, a significant portion of the reactor coolant 
pressure boundary, are important to the safe operation of a pressurized 
water reactor. As such, the NRC has established requirements pertaining 
to the design, fabrication, erection, testing, and inspection of the 
steam generator tubes. With respect to the performance of the PSI of 
steam generator tubes, the NRC has indicated in NRC Regulatory Guide 
(RG) 1.83, Revision 1, ``Inservice Inspection of Pressurized Water 
Reactor Steam Generator Tubes,'' (withdrawn in 2009) that all tubes in 
the steam generator should be inspected by eddy current or alternative 
technique prior to service to establish a baseline condition of the 
tubing. A similar position is articulated in NUREG-0800, Standard 
Review Plan (SRP) Section 5.4.2.2, ``Steam Generator Tube Inservice 
Inspection,'' Revision 1 and subsequent revisions. A PSI is important 
since it ensures that the steam generator tubes are acceptable for 
initial operation. In addition, the PSI provides the baseline condition 
of the tubes. This data is essential in assessing the nature of 
indications found in the tubes during subsequent inservice inspections.
    Preservice requirements for ASME Class 1 components are provided in 
IWB-2200, and IWB-2200(c) currently states, ``Steam generator tube 
examination shall be governed by the plant Technical Specifications 
(TS).'' However, there are no preservice examination requirements for 
steam generators defined in plant TS. Preservice examination 
requirements for steam generators are not within any of the categories 
described in 50.36 for the content of TS. Because IWB-2200(c) requires 
the steam generator tube examinations be performed in accordance with 
plant TS, and TS contain no rules for PSI of steam generator tubing, 
the NRC is clarifying the preservice inspection requirements for steam 
generator tubes.
    The proposed clarification is consistent with industry guidelines 
and the NRC staff position outlined in SRP Section 5.4.2.2, ``Steam 
Generator Program.'' The proposed requirement supersedes the 
requirements of IWB-2200(c). These inspections must be performed with 
the objective of finding and characterizing the types of preservice 
flaws that may be present in the tubes and flaws that may occur during 
operation.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping 
Devices
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxi) to prohibit the 
use of mechanical clamping devices on Class 1 piping and portions of 
piping systems that form the containment boundary. In the 2010 Edition 
of the ASME BPV Code, a change was made to include mechanical clamping 
devices under the small items exclusion rules of IWA-4131. Currently in 
the 2007 Edition/2008 Addenda of Section XI under IWA-4133, 
``Mechanical Clamping Devices Used as Piping Pressure Boundary,'' 
mechanical clamping devices may be used only if they meet the 
requirements of Mandatory Appendix IX of Section XI of the ASME BPV 
Code. Article IX-1000 (c) of Appendix IX prohibits the use of 
mechanical clamping devices on (1) Class 1 piping and (2) portions of a 
piping system that form the containment boundary.
    In the 2010 Edition, IWA-4133 was modified to allow use of IWA-
4131.1(c) for the installation of mechanical clamping devices. This 
change allowed

[[Page 56828]]

the use of small items exemption rules in the installation of 
mechanical clamps. Subparagraph IWA-4131.1(c) was added such that 
mechanical clamping devices installed on items classified as ``small 
items'' under IWA-4131, including Class 1 piping and portions of a 
piping system that form the containment boundary, would be allowed 
without a repair/replacement plan, pressure testing, services of an 
Authorized Inspection Agency, and completion of NIS-2 form.
    The NRC, in accordance with the previously approved IWA-4133 of the 
2007 Edition/2008 Addenda of the ASME BPV Code, does not believe that 
the ASME has provided a sufficient technical basis to support the use 
of mechanical clamps on Class 1 piping or portions of a piping system 
that form the containment boundary as a permanent repair. Furthermore, 
the NRC does not believe that the ASME has provided any basis for the 
small item exemption allowing the installation of mechanical clamps on 
these components. In the 2011 Addenda of the ASME BPV Code, IWA-
4131.1(c) was relocated to IWA-4131.1(d).
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report 
Submittal
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxii) to require 
licensees using the 2010 Edition and later editions and addenda of 
Section XI to continue to submit Summary Reports as required in IWA-
6240 of the 2009 Addenda.
    Prior to the 2010 Edition, Section XI required the preservice 
summary report to be submitted prior to the date of placement of the 
unit into commercial service, and the inservice summary report to be 
submitted within 90 calendar days of the completion of each refueling 
outage. In the 2010 Edition, IWA-6240 was revised to state, ``Summary 
Reports shall be submitted to the enforcement and regulatory 
authorities having jurisdiction at the plant site, if required by these 
authorities.'' This change in the 2010 Edition could lead to confusion 
as to whether or not the summary reports need to be submitted to the 
NRC, as well as the time for submitting the reports if they were 
required. The NRC believes that summary reports must continue to be 
submitted to the NRC in a timely manner because they provide valuable 
information regarding examinations performed, conditions noted, 
corrective actions taken, and the implementation status of PSI and ISI 
programs. Therefore, the NRC proposes adding Sec.  50.55a(b)(2)(xxxii) 
to ensure that preservice and inservice summary reports will continue 
to be submitted within the timeframes currently established in Section 
XI editions and addenda prior to the 2010 Edition.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed 
Allowable Pressure
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxiii) to prohibit the 
use of Appendix G Paragraph G-2216 in the 2011 Addenda and later 
editions and addenda of the ASME BPV Code, Section XI. The 2011 Addenda 
of the ASME BPV Code included, for the first time, a risk-informed 
methodology to compute allowable pressure as a function of inlet 
temperature for reactor heat-up and cool-down at rates not to exceed 
100 degrees F/hr (56 degrees C/hr). This methodology was developed 
based upon probabilistic fracture mechanics (PFM) evaluations that 
investigated the likelihood of reactor pressure vessel (RPV) failure 
based on specific heat-up and cool-down scenarios.
    During the ASME's consideration of this change, the NRC staff noted 
that additional requirements would need to be placed on the use of this 
alternative. For example, the NRC staff indicated that it would be 
important for a licensee who wishes to utilize such a risk-informed 
methodology for determining plant-specific pressure-temperature limits 
to ensure that the material condition of its facility is consistent 
with assumptions made in the PFM evaluations that supported the 
development of the methodology. One aspect of this would be evaluating 
plant-specific inservice inspection data to determine whether the 
facility's RPV flaw distribution was consistent with the flaw 
distribution assumed in the supporting PFM evaluations. This 
consideration is consistent with a similar requirement established by 
the NRC in Sec.  50.61a, ``Alternative Fracture Toughness Requirements 
for Protection against Pressurized Thermal Shock Events.'' The PFM 
methodology that supports Sec.  50.61a is very similar that which was 
used to support ASME BPV Code, Section XI, Appendix G, Paragraph G-
2216. These concerns with the Paragraph G-2216 methodology for 
computing allowable pressure as a function of inlet temperature for 
reactor heat-up and cooldown were not addressed by the ASME. 
Accordingly, the NRC is proposing to prohibit the use of Paragraph G-
2216 in Appendix G of the 2010 Edition. The continued use of the 
deterministic methodology of Section XI, Appendix G to generate P-T 
limits remains acceptable.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Disposition of Flaws 
in Class 3 Components
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxiv) to require that 
when using the 2013 Edition of the ASME BPV Code, Section XI, the 
licensee shall use the acceptance standards of IWD-3510 for the 
disposition of flaws in Category D-A components (i.e., welded 
attachments for vessels, piping, pumps, and valves).
    The 2013 Edition of the ASME BPV Code, Section XI, IWD-3510, 
``Standards for Examination Category D-A, Welded Attachments for 
Vessels, Piping, Pumps, and Valves,'' states that the acceptance 
standards are: ``In the course of preparation, the requirements of IWC-
3500 may be used.'' The ASME BPV Code, Section XI, IWD-3410, 
``Acceptance Standards,'' states that the acceptance standards 
referenced in Table IWD-3410-1 shall be applied to determine 
acceptability for service. Table IWD-3410-1 states that the acceptance 
standard for Examination Category D-A is IWB-3510.
    A discrepancy exists between the provisions in IWD-3410, which 
references Table IWD-3410-1, and the provisions in IWD-3510. The 
provisions in IWD-3510 require the use of the acceptance standards of 
IWC-3500 whereas Table IWD-3410-1 requires the use of the acceptance 
standards of IWB-3510 to disposition flaws detected in the Examination 
Category D-A components. Both IWD-3410 and IWD-3510 should reference 
the same subarticle and acceptance standards. The NRC believes that 
this discrepancy is due to an error in the publishing of the 2013 
Edition because the code committee action which proposed the revised 
Class 3 acceptance criteria and added Table IWD-3410-1 showed the 
appropriate Acceptance Standard to be IWD-3510. The intent of the 
condition is to provide clarification and consistency in requirements 
between IWD-3410 and IWD-3510.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 
in the KIa and KIc Equations
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxv) to specify that 
when licensees use the 2013 Edition of the ASME BPV Code, Section XI, 
Appendix A, paragraph A-4200, if T0 is available, then 
RTT0 may be used in place of RTNDT for 
applications using the KIc equation and the associated 
KIc curve, but not for applications using the KIa 
equation and the associated KIa curve.
    Non-mandatory Appendix A provides a procedure based on linear 
elastic

[[Page 56829]]

fracture mechanics (LEFM) for determining the acceptability of flaws 
that have been detected during inservice inspections that exceed the 
allowable flaw indication standards of IWB-3500. Sub-article A-4200 
provides a procedure for determining fracture toughness of the material 
used in the LEFM analysis. The NRC staff's concern is related to the 
proposed insertion regarding an alternative based on Master Curve 
methodology to determine the nil-ductility transition reference 
temperature RTNDT, which is an important parameter in 
determining the fracture toughness of the material. Specifically, the 
insertion proposed to use Master Curve reference temperature 
RTT0, which is defined as RTT0 = T0 + 
35 [deg]F, where T0 is a material-specific temperature value 
determined in accordance with ASTM E1921, ``Standard Test Method for 
Determination of Reference Temperature, T0, for Ferritic 
Steels in the Transition Range,'' to index (shift) the fracture 
toughness KIc curve, based on the lower bound of static 
initiation critical stress intensity factor, as well as the 
KIa curve, based on the lower bound of crack arrest critical 
stress intensity factor.
    While use of RTT0 to index the KIc curve is 
acceptable, using RTT0 to index the KIa curve is 
questionable. This NRC staff concern is based on the data analysis in 
``A Physics-Based Model for the Crack Arrest Toughness of Ferritic 
Steels,'' written by NRC staff member Mark Kirk, and published in 
``Fatigue and Fracture Mechanics, 33rd Volume, ASTM STP 1417,'' which 
indicated that the crack arrest data does not support using 
RTT0 as RTNDT to index the KIa curve. 
This is also confirmed by industry data disclosed in a presentation, 
``Final Results from the CARINA Project on Crack Initiation and Arrest 
of Irradiated German RPV Steels for Neutron Fluences in the Upper 
Bound,'' by AREVA at the 26th Symposium on Effects of Radiation on 
Nuclear Materials (June 12-13, 2013, Indianapolis, IN, USA). The NRC 
staff recognized that the proposed insertion is consistent with Code 
Case N-629, ``Use of Fracture Toughness Test Data to Establish 
Reference Temperature for Pressure Retaining Materials,'' which was 
accepted by the NRC without conditions. In addition to the current NRC 
effort, the appropriate ASME Code committee is in the process of 
correcting this issue in a future revision of Appendix A of Section XI.
    With this condition, users of Appendix A can avoid using an 
erroneous fracture toughness KIa value in their LEFM 
analysis for determining the acceptability of a detected flaw in 
applicable components. Therefore, the NRC is proposing to add a 
condition which permits the use of RTT0 in place of 
RTNDT in applications using the KIc equation and 
the associated KIc curve, but does not permit the use of 
RTT0 in place of RTNDT in applications using the 
KIa equation and the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of 
Irradiated Materials
    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvi) to require 
licensees using ASME BPV Code, Section XI, 2013 Edition, Appendix A, 
paragraph A-4400, to obtain NRC approval before using irradiated 
T0 and the associated RTT0 in establishing 
fracture toughness of irradiated materials.
    Sub-article A-4400 provides guidance for considering irradiation 
effects on materials. The NRC staff's concern is related to use of 
RTT0 based on measured T0 of the irradiated 
materials. Specifically, the NRC staff has concerns over this sentence 
in the proposed insertion: ``Measurement of RTT0 of 
unirradiated or irradiated materials as defined in A-4200(b) is 
permitted, including use of the procedures given in ASTM E1921 to 
obtain direct measurement of irradiated T0.''
    Permission of measurement of RTT0 of irradiated 
materials, without providing guidelines regarding how to use the 
measured parameter in determining the fracture toughness of the 
irradiated materials, may mislead the users of Appendix A into adopting 
methodology not accepted by the NRC. With this condition, users of 
Appendix A can avoid using a fracture toughness KIc value 
based on the irradiated T0 and the associated 
RTT0 in their LEFM analysis for determining the 
acceptability of a detected flaw in applicable components.
10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements
    The NRC proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and 
(g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i), 
(g)(3)(ii), and (g)(3)(v) to distinguish the requirements for 
accessibility and preservice examination from those for inservice 
inspection in Sec.  50.55a(g). No substantive change to the 
requirements is intended by these revisions.

C. ASME OM Code

10 CFR 50.55a(b)(3) Conditions on ASME OM Code
    The NRC proposes to revise Sec.  50.55a(b)(3) to clarify that 
Subsections ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory 
Appendices I, II, III, and V; and Non-mandatory Appendices A through H 
and J through M of the ASME OM Code would be incorporated by reference 
in Sec.  50.55a. The NRC is clarifying that the ASME OM Code non-
mandatory appendices, which are incorporated by reference into Sec.  
50.55a are approved for use, but are not mandated. The non-mandatory 
appendices may be used by applicants and licensees of nuclear power 
plants, subject to the conditions in Sec.  50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
    The NRC proposes to revise Sec.  50.55a(b)(3)(i) to allow use of 
the 1983 Edition through the 1994 Edition, 2008 Edition, and the 2009-
1a Addenda of NQA-1, ``Quality Assurance Requirements for Nuclear 
Facility Applications.'' The NRC reviewed these Editions and Addenda 
after the 1983 Edition and compared them to the previously approved 
versions of NQA-1 and found that there were no significant differences.
    The NRC is considering removing the reference in Sec.  
50.55a(b)(3)(i) to versions of NQA-1 older than the 1994 Edition. The 
NRC requests public comment on whether any licensee is committed to, 
and is using, a version of NQA-1 older than the 1994 Edition and, if 
so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) 
Testing
    The NRC proposes to revise Sec.  50.55a(b)(3)(ii) to reflect the 
new Appendix III, ``Preservice and Inservice Testing of Active Electric 
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,'' 
of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. 
Appendix III of the ASME OM Code establishes provisions for periodic 
verification of the design-basis capability of MOVs within the scope of 
the IST program. Appendix III of the ASME OM Code reflects the 
incorporation of ASME OM Code Cases OMN-1, ``Alternative Rules for 
Preservice and Inservice Testing of Active Electric Motor-Operated 
Valve Assemblies in Light-Water Reactor Power Plants,'' and OMN-11, 
``Risk-Informed Testing for Motor-Operated Valves.'' The NRC proposes 
to add four conditions in new Sec. Sec.  50.55a(b)(3)(ii)(A), (B), (C), 
and (D) to address periodic verification of MOV design-basis 
capability. These conditions are discussed in the next four sections.

[[Page 56830]]

10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval
    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(A) to require that 
licensees evaluate the adequacy of the diagnostic test interval for 
each MOV and adjust the interval as necessary, but not later than 5 
years or three refueling outages (whichever is longer) from initial 
implementation of ASME OM Code, Appendix III. Paragraph III-3310(b) in 
Appendix III includes a provision stating that if insufficient data 
exist to determine the IST interval, then MOV inservice testing shall 
be conducted every two refueling outages or 3 years (whichever is 
longer) until sufficient data exist, from an applicable MOV or MOV 
group, to justify a longer IST interval. As discussed in 64 FR 51386 
(September 22, 1999) with respect to the use of ASME OM Code Case OMN-
1, the NRC considers it appropriate to include a modification requiring 
licensees to evaluate the information obtained for each MOV, during the 
first 5 years or three refueling outages (whichever is longer) of the 
use of Appendix III to validate assumptions made in justifying a longer 
test interval.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact On Risk
    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(B) to require that 
licensees ensure that the potential increase in core damage frequency 
(CDF) and large early release frequency (LERF) associated with the 
extension is acceptably small when extending exercise test intervals 
for high risk MOVs beyond a quarterly frequency. As discussed in 64 FR 
51386 (September 22, 1999) with respect to the use of ASME OM Code Case 
OMN-1, the NRC considers it important for licensees to have sufficient 
information from the specific MOV, or similar MOVs, to demonstrate that 
exercising on a refueling outage frequency does not significantly 
affect component performance. The information may be obtained by 
grouping similar MOVs and establishing periodic exercising intervals of 
MOVs in the group over the refueling interval.
    Section 50.55a(b)(3)(ii)(B) requires that the increase in the 
overall plant CDF and LERF resulting from the extension be acceptably 
small. As presented in RG 1.174, ``An Approach for Using Probabilistic 
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 
the Licensing Basis,'' the NRC considers acceptably small changes to be 
relative and to depend on the current plant CDF and LERF. For plants 
with total baseline CDF of 10^-4 per year or less, acceptably 
small means CDF increases of up to 10^-5 per year and for 
plants with total baseline CDF greater than 10^-4 per year, 
acceptably small means CDF increases of up to 10^-6 per year. 
For plants with total baseline LERF of 10^-5 per year or 
less, acceptably small LERF increases are considered to be up to 
10^-6 per year, and for plants with total baseline LERF 
greater than 10^-5 per year, acceptably small LERF increases 
are considered to be up to 10^-7 per year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization
    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(C) to require, when 
applying Appendix III to the ASME OM Code, that licensees categorize 
MOVs according to their safety significance using the methodology 
described in ASME OM Code Case OMN-3, ``Requirements for Safety 
Significance Categorization of Components Using Risk Insights for 
Inservice Testing of LWR Power Plants,'' subject to the conditions 
discussed in RG 1.192, or using an MOV risk ranking methodology 
accepted by the NRC on a plant-specific or industry-wide basis in 
accordance with the conditions in the applicable safety evaluation. 
Paragraph III-3720 in Appendix III to the ASME OM Code states that when 
applying risk insights, each MOV shall be evaluated and categorized 
using a documented risk ranking methodology. Further, Appendix III only 
addresses risk ranking methodologies that include two risk categories. 
In light of the potential extension of quarterly test intervals for 
high risk MOVs and the relaxation of IST activities for low risk MOVs 
based on risk insights, the NRC has determined that the rule should 
specify that risk ranking methodologies must have been accepted by the 
NRC through RG 1.192 (which accepts ASME OM Code Case OMN-3 with the 
specified conditions) or safety evaluations issued to address plant-
specific or industry-wide risk ranking methodologies.
    Two conditions that were previously in RG 1.192 on the use of ASME 
OM Code Case OMN-11 related to application of the test interval 
criteria and grouping for low safety significant MOVs have been 
incorporated in an acceptable manner in Appendix III to the ASME OM 
Code. As noted in RG 1.192 on the use of ASME OM Code Case OMN-1, the 
benefits of performing a particular test should be balanced against the 
potential adverse effects placed on the valves or systems caused by 
this testing.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time
    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(D) to require that 
when a licensee applies Paragraph III-3600, ``MOV Exercising 
Requirements,'' of Appendix III to the OM Code, the licensee verify 
that the stroke time of the MOV satisfies the assumptions in the plant 
safety analyses. Previous editions and addenda of the ASME OM Code 
specified that the licensee must perform quarterly MOV stroke time 
measurements that could be used to verify that the MOV stroke time 
satisfies the assumptions in the safety analyses consistent with plant 
TS. The need for verification of the MOV stroke time during periodic 
exercising is consistent with the NRC's lessons learned from the 
implementation of ASME OM Code Case OMN-1. However, Paragraph III-3600 
of Appendix III of the versions of the OM Code proposed to be 
incorporated by reference in this rulemaking no longer require the 
verification of MOV stroke time during periodic exercising. For this 
reason, the NRC is proposing to adopt the new condition which will 
effectively retain the need to verify MOV stroke time during periodic 
exercising.
10 CFR 50.55a(b)(3)(iii) OM condition: New Reactors
    The NRC proposes to add Sec.  50.55a(b)(3)(iii) to apply specific 
conditions for IST programs applicable to licensees of new nuclear 
power plants in addition to the provisions of the ASME OM Code as 
incorporated by reference with conditions in Sec.  50.55a. Licensees of 
``new reactors'' are, as identified in the proposed paragraph: (i) 
Holders of operating licenses for nuclear power reactors that received 
construction permits under this part on or after the date 12 months 
after the effective date of this rulemaking and (ii) holders of 
combined licenses (COLs) issued under 10 CFR part 52, whose initial 
fuel loading occurs on or after the date 12 months after the effective 
date of this rulemaking. This implementation schedule for new reactors 
is consistent with the NRC regulations in Sec.  50.55a(f)(4)(i).
    The NRC is evaluating COL applications to construct and operate 
nuclear power plants with certified designs under the process described 
in 10 CFR part 52. Commission Papers SECY-90-016, ``Evolutionary Light 
Water Reactor (LWR) Certification Issues and Their Relationship to 
Current Regulatory Requirements;'' SECY-93-087, ``Policy, Technical, 
and Licensing Issues Pertaining to Evolutionary and

[[Page 56831]]

Advanced Light-Water Reactor (ALWR) Designs;'' SECY-94-084, ``Policy 
and Technical Issues Associated with the Regulatory Treatment of Non-
Safety Systems (RTNSS) in Passive Plant Designs;'' and SECY-95-132, 
``Policy and Technical Issues Associated with the Regulatory Treatment 
of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY-94-084),'' 
discuss IST programs for new reactors licensed under 10 CFR part 52.
    In recognition of new reactor designs and lessons learned from 
nuclear power plant operating experience, the ASME is updating the OM 
Code to incorporate improved IST provisions for components used in 
nuclear power plants that were issued (or will be issued) construction 
permits, or COLs, on or following January 1, 2000 (defined in the ASME 
OM Code as post-2000 plants). The first phase of the ASME effort 
incorporated IST provisions that specify full flow pump testing and 
other clarifications for post-2000 plants in the ASME OM Code beginning 
with the 2011 Addenda. The second phase of the ASME effort incorporated 
preservice and inservice inspection and surveillance provisions for 
pyrotechnic-actuated (squib) valves in the 2012 Edition of the ASME OM 
Code. The ASME is considering further modifications to the ASME OM Code 
to address additional lessons learned from valve operating experience 
and new reactor issues. As described in the following paragraphs, Sec.  
50.55a(b)(3)(iii) will include four specific conditions.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves
    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(A) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) develop a program to 
periodically verify the capability of power-operated valves (POVs) to 
perform their design-basis safety functions. While Appendix III to the 
ASME OM Code addresses this requirement for motor-operated valves 
(MOVs) with applicable conditions specified in Sec.  50.55a, nuclear 
power plant licensees will need to develop programs to periodically 
verify the design-basis capability of other POVs. The NRC's Regulatory 
Issue Summary (RIS) 2000-03, ``Resolution of Generic Issue 158: 
Performance of Safety-Related Power-Operated Valves Under Design Basis 
Conditions,'' provides attributes for a successful long-term periodic 
verification program for POVs by incorporating lessons learned from MOV 
performance at operating nuclear power plants and during research 
programs. Implementation of Appendix III to the ASME OM Code as 
accepted in Sec.  50.55a(b)(3)(ii) is acceptable in satisfying Sec.  
50.55a(b)(3)(iii)(A) for MOVs.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(B) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) perform bi-directional 
testing of check valves within the IST program where practicable. 
Nuclear power plant operating experience has revealed that testing 
check valves in only the flow direction can result in significant 
degradation, such as a missing valve disc, not being identified by the 
test. Nonmandatory Appendix M, ``Design Guidance for Nuclear Power 
Plant Systems and Component Testing,'' to ASME OM Code, 2011 Addenda 
and 2012 Edition, includes guidance for the design of new reactors to 
enable bi-directional testing of check valves. New reactor designs will 
provide the capability for licensees of new nuclear power plants to 
perform bi-directional testing of check valves within the IST program.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration
    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(C) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) monitor flow-induced 
vibration (FIV) from hydrodynamic loads and acoustic resonance during 
preservice testing and inservice testing to identify potential adverse 
flow effects that might impact components within the scope of the IST 
program. Nuclear power plant operating experience has revealed the 
potential for adverse flow effects from vibration caused by 
hydrodynamic loads and acoustic resonance on components in the reactor 
coolant, steam, and feedwater systems. Therefore, the licensee will 
need to address potential adverse flow effects on safety-related pumps, 
valves, and dynamic restraints within the IST program in the reactor 
coolant, steam, and feedwater systems from hydraulic loading and 
acoustic resonance during plant operation to confirm that piping, 
components, restraints, and supports have been designed to withstand 
the dynamic effects of steady-state FIV and anticipated operational 
transient conditions. The initial test program can be used to verify 
that safety-related piping and components are properly installed and 
supported such that vibrations caused by steady-state or dynamic 
effects do not result in excessive stress or fatigue in safety-related 
plant systems.
10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems
    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(D) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) establish a program to 
assess the operational readiness of pumps, valves, and dynamic 
restraints within the scope of the Regulatory Treatment of Non-Safety 
Systems (RTNSS) for applicable reactor designs. In SECY-94-084 and 
SECY-95-132, the Commission discusses RTNSS policy and technical issues 
associated with passive plant designs. Some new nuclear power plants 
have ALWR designs that use passive safety systems that rely on natural 
forces, such as density differences, gravity, and stored energy, to 
supply safety injection water and to provide reactor core and 
containment cooling. Active systems in passive ALWR designs are 
categorized as non-safety systems with limited exceptions. Active 
systems in passive ALWR designs provide the first line of defense to 
reduce challenges to the passive systems in the event of a transient at 
the nuclear power plant. Active systems that provide a defense-in-depth 
function in passive ALWR designs need not meet all of the acceptance 
criteria for safety-related systems. However, there should be a high 
level of confidence that these active systems will be available and 
reliable when challenged. The combined activities to provide confidence 
in the capability of these active systems in passive ALWR designs to 
perform their functions important to safety are referred to together as 
the RTNSS program. In a public memorandum dated July 24, 1995, the NRC 
staff provided a consolidated list of the approved policy and technical 
positions associated with RTNSS equipment in passive plant designs 
discussed in SECY-94-084 and SECY-95-132 (ADAMS Accession No. 
ML003708048). This new paragraph will specify the need for licensees to 
assess the operational readiness of RTNSS pumps, valves, and dynamic 
restraints.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
    The NRC proposes to revise Sec.  50.55a(b)(3)(iv) to address 
Appendix II, ``Check Valve Condition Monitoring Program,'' provided in 
the 2003 Addenda through the 2012 Edition of the ASME OM Code. In the 
2003 Addenda of the ASME OM Code, ASME revised Appendix II to address 
the conditions specified in Sec.  50.55a for older versions of the 
appendix. Therefore, the NRC considers Appendix

[[Page 56832]]

II in the 2003 Addenda through the 2012 Edition of the ASME OM Code to 
be acceptable for use without conditions. In accepting the recent 
versions of Appendix II, the NRC proposes to clarify that (1) the 
maximum test interval allowed by Appendix II for individual check 
valves in a group of two valves or more must be supported by periodic 
testing of a sample of check valves in the group during the allowed 
interval and (2) the periodic testing plan must be designed to test 
each valve of a group at approximate equal intervals not to exceed the 
maximum requirement interval. The NRC notes that ASME has provided 
additional improvements to Appendix II since issuance of the 2003 
Addenda. Therefore, where a licensee plans to voluntarily implement 
Appendix II to the ASME OM Code, the licensee should apply Appendix II 
in the most recent addenda and edition of ASME OM Code incorporated by 
reference in Sec.  50.55a. The conditions currently specified for the 
use of Appendix II, 1995 Edition with the 1996 and 1997 Addenda, and 
1998 Edition through the 2002 Addenda, of the OM Code remain the same 
in this proposed rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
    The NRC proposes to add Sec.  50.55a(b)(3)(vii) to prohibit the use 
of Subsection ISTB, ``Inservice Testing of Pumps in Light-Water Reactor 
Nuclear Power Plants,'' in the 2011 Addenda of the ASME OM Code. In the 
2011 Addenda to the ASME OM Code, the upper end of the Acceptable Range 
and the Required Action Range for flow and differential or discharge 
pressure for comprehensive pump testing in Subsection ISTB was raised 
to higher values. The NRC staff on the ASME OM Code committee accepted 
the proposed increase of the upper end of the Acceptable Range and 
Required Action Range with the planned addition of a requirement for a 
pump periodic verification test program in the ASME OM Code. However, 
the 2011 Addenda to the ASME OM Code did not include the requirement 
for a pump periodic verification test program as an oversight. Since 
then, the 2012 Edition to the ASME OM Code has incorporated Mandatory 
Appendix V, ``Pump Periodic Verification Test Program,'' that supports 
the changes to the acceptable and required action ranges for 
comprehensive pump testing in Subsection ISTB. Therefore, proposed new 
Sec.  50.55a(b)(3)(vii) would prohibit the use of Subsection ISTB in 
the 2011 Addenda of the ASME OM Code. Licensees will be allowed to 
apply Subsection ISTB with the revised acceptable and required action 
ranges in the 2012 Edition of the ASME OM Code as incorporated by 
reference in Sec.  50.55a.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
    The NRC proposes to add Sec.  50.55a(b)(3)(viii) to specify that 
licensees proposing to implement Subsection ISTE, ``Risk-Informed 
Inservice Testing of Components in Light-Water Reactor Nuclear Power 
Plants,'' of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 
Edition, must request and obtain NRC authorization in accordance with 
Sec.  50.55a(z) to apply Subsection ISTE on a plant-specific basis as a 
risk-informed alternative to the applicable IST requirements in the 
ASME OM Code.
    In the 2009 Edition of the ASME OM Code, the ASME included new 
Subsection ISTE that describes a voluntary risk-informed approach in 
developing an IST program for pumps and valves at nuclear power plants. 
If a licensee chooses to implement this risk-informed IST approach, 
Subsection ISTE indicates that all requirements in Subsection ISTA, 
``General Requirements,'' Subsection ISTB, and Subsection ISTC, 
``Inservice Testing of Valves in Light-Water Reactor Nuclear Power 
Plants,'' of the ASME OM Code continue to apply, except those 
identified in Subsection ISTE. The ASME selected risk-informed guidance 
from ASME OM Code Cases OMN-1, OMN-3, OMN-4, ``Requirements for Risk 
Insights for Inservice Testing of Check Valves at LWR Power Plants,'' 
OMN-7, ``Alternative Requirements for Pump Testing,'' OMN-11, and OMN-
12, ``Alternative Requirements for Inservice Testing Using Risk 
Insights for Pneumatically and Hydraulically Operated Valve Assemblies 
in Light-Water Reactor Power Plants,'' in preparing Subsection ISTE of 
the ASME OM Code.
    During development of Subsection ISTE, the NRC staff participating 
on the ASME OM Code committees indicated that the conditions specified 
in RG 1.192 for the use of the applicable ASME OM Code Cases need to be 
considered when evaluating the acceptability of the implementation of 
Subsection ISTE. In addition, the NRC staff noted that several aspects 
of Subsection ISTE will need to be addressed on a case-by-case basis 
when determining the acceptability of its implementation. Therefore, 
new Sec.  50.55a(b)(3)(viii) requires that licensees proposing to 
implement Subsection ISTE of the ASME OM Code must request approval 
from the NRC to apply Subsection ISTE on a plant-specific basis as a 
risk-informed alternative to the applicable IST requirements in the 
ASME OM Code.
    Nuclear power plant applicants for construction permits under 10 
CFR part 50, or combined licenses for construction and operation under 
10 CFR part 52, may describe their proposed implementation of the risk-
informed IST approach specified in Subsection ISTE of the ASME OM Code 
for NRC review in their applications.
    The NRC will evaluate Sec.  50.55a(z) requests for approval to 
implement Subsection ISTE in accordance with the following 
considerations:
1. Scope of Risk-Informed IST Program
    Subsection ISTE-1100, ``Applicability,'' establishes the component 
safety categorization methodology and process for dividing the 
population of pumps and valves, as identified in the IST Program Plan, 
into high safety significant component (HSSC) and low safety 
significant component (LSSC) categories. When establishing a risk-
informed IST program, the licensee should address a wide range of 
components important to safety at the nuclear power plant that includes 
both safety-related and nonsafety-related components. These components 
might extend beyond the scope of the ASME OM Code.
2. Risk-Ranking Methodology
    The licensee should specify in its request for authorization to 
implement a risk-informed IST program the methodology to be applied in 
risk ranking its components. ISTE-4000, ``Specific Component 
Categorization Requirements,'' incorporates ASME OM Code Case OMN-3 for 
the categorization of pumps and valves in developing a risk-informed 
IST program. The OMN-3 Code Case methodology for risk ranking uses two 
categories of safety significance. The NRC staff has also accepted 
other methodologies for risk ranking that use three categories of 
safety significance.
3. Safety Significance Categorization
    The licensee should categorize components according to their safety 
significance based on the methodology described in Subsection ISTE with 
the applicable conditions on the use of ASME OM Code Case OMN-3 
specified in RG 1.192, or use other risk ranking methodologies accepted 
by the NRC on a plant-specific or industry-wide basis with applicable 
conditions specified by the NRC for their acceptance. The licensee 
should address the seven

[[Page 56833]]

conditions in RG 1.192 for the use of ASME OM Code Case OMN-3 as 
appropriate in developing the risk-informed IST program described in 
Subsection ISTE. With respect to the provisions in Subsection ISTE, 
these conditions are:
    (a) The implementation of ISTE-1100 should include within the scope 
of a licensee's risk-informed IST program non-ASME Code pumps and 
valves categorized as HSSCs that might not currently be included in the 
IST program at the nuclear power plant.
    (b) The decision criteria discussed in ISTE-4410, ``Decision 
Criteria,'' and Non-mandatory Appendix L, ``Acceptance Guidelines,'' of 
the ASME OM Code for evaluating the acceptability of aggregate risk 
effects (i.e., for Core Damage Frequency [CDF] and Large Early Release 
Frequency [LERF]) should be consistent with the guidance provided in RG 
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis.''
    (c) The implementation of ISTE-4440, ``Defense in Depth,'' should 
be consistent with the guidance contained in Section 2.2.1, ``Defense-
in-Depth Evaluation,'' and Section 2.2.2, ``Safety Margin Evaluation,'' 
of RG 1.175, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Inservice Testing.''
    (d) The implementation of ISTE-4500, ``Inservice Testing Program,'' 
and ISTE-6100, ``Performance Monitoring,'' should be consistent with 
the guidance contained in Section 3.2, ``Program Implementation,'' and 
Section 3.3, ``Performance Monitoring,'' of RG 1.175.
    (e) The implementation of ISTE-3210, ``Plant-Specific PRA,'' should 
be consistent with the guidance that the Owner is responsible for 
demonstrating and justifying the technical adequacy of the PRA analyses 
used as the basis to perform component risk ranking and for estimating 
the aggregate risk impact. For example, RG 1.200, ``An Approach for 
Determining the Technical Adequacy of Probabilistic Risk Assessment 
Results for Risk-Informed Activities,'' and RG 1.201, ``Guidelines for 
Categorizing Structures, Systems, and Components in Nuclear Power 
Plants According to their Safety Significance,'' provide guidance for 
PRA technical adequacy and component risk ranking.
    (f) The implementation of ISTE-4240, ``Reconciliation,'' should 
specify that the expert panel may not classify components that are 
ranked HSSC by the results of a qualitative or quantitative PRA 
evaluation (excluding the sensitivity studies) or the defense-in-depth 
assessment to LSSC.
    (g) The implementation of ISTE-3220, ``Living PRA,'' should be 
consistent with the following: (i) To account for potential changes in 
failure rates and other changes that could affect the PRA, changes to 
the plant must be reviewed and, as appropriate, the PRA updated; (ii) 
when the PRA is updated, the categorization of structures, systems, and 
components must be reviewed and changed if necessary to remain 
consistent with the categorization process; and (iii) the review of the 
plant changes must be performed in a timely manner and must be 
performed once every two refueling outages, or as required by Sec.  
50.71(h)(2) for COL holders.
4. Pump Testing
    Subsection ISTE-5100, ``Pumps,'' incorporates ASME OM Code Case 
OMN-7 for risk-informed testing of pumps categorized as LSSCs. 
Subsection ISTE-5100 allows the interval for Group A and Group B 
testing of LSSC pumps specified in Subsection ISTB of the ASME OM Code 
to be extended from the current 3-month interval to intervals of 6 
months or 2 years. Subsection ISTE-5100 eliminates the requirement in 
Subsection ISTB to perform comprehensive pump testing for LSSC pumps. 
Table ISTE-5121-1, ``LSSC Pump Testing,'' specifies that pump operation 
may be required more frequently than the specified test frequency (6 
months) to meet vendor recommendations. Subsection ISTE-4500, 
``Inservice Testing Program,'' specifies in ISTE-4510, ``Maximum 
Testing Interval,'' that the maximum testing interval shall be based on 
the more limiting of (a) the results of the aggregate risk, or (b) the 
performance history of the component. ISTE-5130, ``Maximum Test 
Interval--Pre-2000 Plants,'' specifies that the most limiting interval 
for LSSC pump testing shall be determined from ISTE-4510 and ISTE-5120, 
``Low Safety Significant Pump Testing.'' The ASME developed the 
comprehensive pump test requirements in the ASME OM Code to address 
weaknesses in the Code requirements to assess the operational readiness 
of pumps to perform their design-basis safety function. Therefore, the 
licensee should ensure that testing under Subsection ISTE will provide 
assurance of the operational readiness of pumps in each safety 
significant categorization to perform their design-basis safety 
function as described in RGs 1.174 and 1.175.
5. Motor-Operated Valve Testing
    Subsection ISTE-5300, ``Motor Operated Valve Assemblies,'' provides 
a risk-informed IST approach instead of the IST requirements for MOVs 
in Mandatory Appendix III to the ASME OM Code. The ASME prepared 
Appendix III to the OM Code to replace the requirement for quarterly 
stroke-time testing of MOVs with a program of periodic exercising and 
diagnostic testing to address lessons learned from nuclear power plant 
operating experience and industry and regulatory research programs for 
MOV performance. Subsection ISTC of the ASME OM Code specifies the 
implementation of Appendix III for periodic exercising and diagnostic 
testing of MOVs to replace quarterly stroke-time testing previously 
required for MOVs. Appendix III incorporates provisions that allow a 
risk-informed IST approach for MOVs as described in ASME OM Code Cases 
OMN-1 and OMN-11. Subsection ISTE-5300 is not consistent with the 
provisions for the risk-informed IST program for MOVs specified in 
Appendix III to the ASME OM Code (and Code Cases OMN-1 and 11). 
Therefore, licensees proposing to implement Subsection ISTE should 
address the provisions in paragraph III-3700, ``Risk-Informed MOV 
Inservice Testing,'' of Appendix III to the ASME OM Code as 
incorporated by reference in Sec.  50.55a with the applicable 
conditions instead of ISTE-5300.
6. Pneumatically and Hydraulically Operated Valve Testing
    Subsection ISTE-5400, ``Pneumatically and Hydraulically Operated 
Valves,'' specifies that licensees test their AOVs and HOVs in 
accordance with Appendix IV to the ASME OM Code. Subsection ISTE-5400 
indicates that Appendix IV is in the course of preparation. The NRC 
staff will need to review Appendix IV prior to accepting its use as 
part of Subsection ISTE. Therefore, licensees proposing to implement 
Subsection ISTE should describe the planned IST provisions for AOVs and 
HOVs in its request for authorization to implement Subsection ISTE.
7. Pump Periodic Verification Test
    Subsection ISTE does not include a requirement to implement the 
pump periodic verification test program specified in Mandatory Appendix 
V to the ASME OM Code, 2012 Edition. The licensee should address the 
consideration of a pump periodic verification test program in its risk-
informed IST program proposed as part

[[Page 56834]]

of the authorization request to implement Subsection ISTE.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
    The NRC proposes to add Sec.  50.55a(b)(3)(ix) for two purposes. 
First, the proposed condition specifies that licensees applying 
Subsection ISTF, ``Inservice Testing of Pumps in Light-Water Reactor 
Nuclear Power Plants--Post-2000 Plants,'' in the 2012 Edition of the OM 
Code shall satisfy the requirements of Mandatory Appendix V, ``Pump 
Periodic Verification Test Program,'' of the OM Code, 2012 Edition. The 
proposed condition also states that Subsection ISTF, 2011 Addenda, is 
not acceptable for use. As previously discussed regarding new Sec.  
50.55a(b)(3)(vii), the upper end of the Acceptable Range and the 
Required Action Range for flow and differential or discharge pressure 
for comprehensive pump testing in Subsection ISTB in the ASME OM Code 
was raised to higher values in combination with the incorporation of 
Mandatory Appendix V, ``Pump Periodic Verification Test Program.'' 
However, Subsection ISTF in the 2011 Addenda and 2012 Edition to the 
ASME OM Code does not include a requirement for a pump periodic 
verification test program. Therefore, new Sec.  50.55a(b)(3)(ix) would 
require that the provisions of Appendix V be applied when implementing 
Subsection ISTF of the 2012 Edition of the OM Code to support the 
application of the upper end of the Acceptable Range and the Required 
Action Range for flow and differential or discharge pressure for 
inservice pump testing in Subsection ISTF. The proposed paragraph would 
prohibit the use of Subsection ISTF in the 2011 Addenda of the OM Code, 
which does not include Appendix V.
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
    The NRC proposes to add a new paragraph, Sec.  50.55a(b)(3)(xi), 
containing a new condition that would specify that when implementing 
ASME OM Code, Subsection ISTC-3700, ``Position Verification Testing,'' 
licensees shall supplement the ASME OM Code provisions as necessary to 
verify that valve operation is accurately indicated. Subsection ISTC-
3700 of the ASME OM Code requires that valves with remote position 
indicators shall be observed locally at least once every 2 years to 
verify that valve operation is accurately indicated. Subsection ISTC-
3700 states that where practicable, this local observation should be 
supplemented by other indications such as the use of flow meters or 
other suitable instrumentation to verify obturator position. Subsection 
ISTC-3700 also states that where local observation is not possible, 
other indications shall be used for verification of valve operation. 
Nuclear power plant operating experience has revealed that reliance on 
indicating lights and stem travel are not sufficient to satisfy the 
requirement in ISTC-3700 to verify that valve operation is accurately 
indicated. Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' to 10 CFR part 50 requires that where generally recognized 
codes and standards are used, they shall be identified and evaluated to 
determine their applicability, adequacy, and sufficiency, and shall be 
supplemented or modified as necessary to assure a quality product in 
keeping with the required safety function. This new condition specifies 
that when implementing ASME OM Code, Subsection ISTC-3700, licensees 
shall develop and implement a method to verify that valve operation is 
accurately indicated by supplementing valve position indicating lights 
with other indications, such as flow meters or other suitable 
instrumentation, to provide assurance of proper obturator position. 
This is not a new requirement but rather a clarification of the intent 
of the existing ASME OM Code. The ASME OM Code specifies obturator 
movement verification in order to detect certain internal valve failure 
modes consistent with the definition of `exercising' found in ISTA-2000 
(i.e., demonstration that the moving parts of a component function). 
Verification of the ability of an obturator to change or maintain 
position is an essential element of valve operational readiness 
determination which is a fundamental aspect of the ASME OM Code. The 
NRC's position is further elaborated in NUREG-1482 Revision 2, 
paragraph 4.2.7.
10 CFR 50.55a(f): Inservice Testing Requirements
    The NRC proposes to revise the introductory text of Sec.  50.55a(f) 
to indicate that systems and components must meet the requirements for 
``preservice and inservice testing'' in the applicable ASME Codes and 
that both activities are referred to as ``inservice testing'' in the 
remainder of paragraph (f). The proposed change clarifies that the ASME 
OM Code includes provisions for preservice testing of components as 
part of its overall provisions for IST programs. No expansion of IST 
program scope is intended by this clarification.
10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision
    The NRC proposes to revise Sec.  50.55a(f)(3)(iii)(A) to ensure 
that the paragraph is applicable to pumps and valves that are within 
the scope of the ASME OM Code. Paragraph ISTA-1100, ``Scope,'' in 
Subsection ISTA, ``General Requirements,'' of the ASME OM Code states 
that the requirements for preservice and inservice testing and 
examination of components in light-water reactor nuclear power plants 
apply to (a) pumps and valves that are required to perform a specific 
function in shutting down a reactor to the safe shutdown condition, in 
maintaining the safe shutdown condition, or in mitigating the 
consequences of an accident; (b) pressure relief devices that protect 
systems or portions of systems that perform one or more of these three 
functions; and (c) dynamic restraints (snubbers) used in systems that 
perform one of more of these three functions, or to ensure the 
integrity of the reactor coolant pressure boundary. This revision will 
align the scope of pumps and valves for inservice testing with the 
scope defined in the ASME OM Code and in SRP Section 3.9.6, 
``Functional Design, Qualification, and Inservice Testing Programs for 
Pumps, Valves, and Dynamic Restraints.''
10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision
    The NRC proposes to revise Sec.  50.55a(f)(3)(iii)(B) to clarify 
that this paragraph is applicable to pumps and valves that are within 
the scope of the ASME OM Code. This revision will align the scope of 
pumps and valves for inservice testing with the scope defined in the 
ASME OM Code and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First 
Provision
    The NRC proposes to revise Sec.  50.55a(f)(3)(iv)(A) to clarify 
that this paragraph is applicable to pumps and valves that are within 
the scope of the ASME OM Code and not covered by paragraph 
(f)(3)(iii)(A) for Class 1 pumps and valves. This revision will align 
the scope of pumps and valves for inservice testing with the scope 
defined in the ASME OM Code and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second 
Provision
    The NRC proposes to revise Sec.  50.55a(f)(3)(iv)(B) to clarify 
that this paragraph is applicable to pumps and valves that are within 
the scope of the ASME OM Code and not covered by paragraph 
(f)(3)(iii)(B) for Class 1 pumps

[[Page 56835]]

and valves. This revision will align the scope of pumps and valves for 
inservice testing with the scope defined in the ASME OM Code and in SRP 
Section 3.9.6.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for 
Operating Plants
    The NRC proposes to revise Sec.  50.55a(f)(4) to clarify that this 
paragraph is applicable to pumps and valves that are within the scope 
of the ASME OM Code. This revision will align the scope of pumps and 
valves for inservice testing with the scope defined in the ASME OM Code 
and in SRP Section 3.9.6.

D. ASME Code Cases

    The NRC proposes to remove the revision number of the three RGs 
currently approved by the Office of the Federal Register for 
incorporation by reference throughout the substantive provisions of 
Sec.  50.55a. The revision numbers for the RGs approved for 
incorporation by reference (currently, RGs 1.84, 1.147, and 1.192) 
would be retained in paragraph (a)(3)(i) through (a)(3)(iii) of Sec.  
50.55a, where the RGs are listed by full title, including revision 
number. These proposed changes would simplify the regulatory language 
containing cross-references to these RGs and reduce the possibility of 
NRC error in preparing future amendments to Sec.  50.55a with respect 
to these RGs. These changes are administrative in nature and do not 
change substantive requirements with respect to the RGs and the Code 
Cases listed in the RGs.
ASME BPV Code Case N-729-4
    On September 10, 2008, the NRC issued a final rule to update Sec.  
50.55a to the 2004 Edition of the ASME Code (73 FR 52730). As part of 
the final rule, Sec.  50.55a(g)(6)(ii)(D) implemented an augmented 
inservice inspection program for the examination of reactor pressure 
vessel (RPV) upper head penetration nozzles and associated partial 
penetration welds. The program required the implementation of ASME BPV 
Code Case N-729-1, with certain conditions.
    The application of ASME BPV Code Case N-729-1 was necessary because 
the inspections required by the 2004 Edition of the ASME BPV Code, 
Section XI were not written to address degradation of the RPV upper 
head penetration nozzles welds by primary water stress corrosion 
cracking (PWSCC). The safety consequences of inadequate inspections can 
be significant. The NRC's determination that the ASME Code required 
inspections are inadequate is based upon operating experience and 
analysis. The absence of an effective inspection regime could, over 
time, result in unacceptable circumferential cracking, or the 
degradation of the RPV upper head or other reactor coolant system 
components by leakage assisted corrosion. These degradation mechanisms 
increase the probability of a loss-of-coolant accident.
    Examination frequencies and methods for RPV upper head penetration 
nozzles and welds are provided in ASME BPV Code Case N-729-1. The use 
of code cases is voluntary, so these provisions were developed, in 
part, with the expectation that the NRC would incorporate the code case 
by reference into the CFR. Therefore, the NRC adopted rule language in 
Sec.  50.55a(g)(6)(ii)(D) requiring implementation of ASME BPV Code 
Case N-729-1, with conditions, in order to enhance the examination 
requirements in the ASME BPV Code, Section XI for RPV upper head 
penetration nozzles and welds. The examinations conducted in accordance 
with ASME BPV Code Case N-729-1 provide reasonable assurance that ASME 
Code allowable limits will not be exceeded and that PWSCC will not lead 
to failure of the RPV upper head penetration nozzles or welds. However, 
the NRC concluded that certain conditions were needed in implementing 
the examinations in ASME BPV Code Case N-729-1. These conditions are 
set forth in Sec.  50.55a(g)(6)(ii)(D).
    On June 22, 2012, the ASME approved the fourth revision of ASME BPV 
Code Case N-729, (N-729-4). This revision changed certain requirements 
based on a consensus review of inspection techniques and frequencies. 
These changes were deemed necessary by the ASME to supersede the 
previous requirements under N-729-1 to establish an effective long-term 
inspection program for the RPV upper head penetration nozzles and 
associated welds in pressurized water reactors. The major changes 
included incorporation of previous NRC conditions in the CFR. Minor 
changes were also made to address editorial issues, to correct figures 
or to add clarity.
    The NRC proposes to update the requirements of Sec.  
50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV Code 
Case N-729-4, with conditions. The NRC conditions have been modified to 
address the changes in ASME BPV Code Case N-729-4. The NRC's proposed 
revisions to the conditions on ASME BPV Code Case N-729-1 are discussed 
in the next four sections.

10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(1) to change 
the version of ASME BPV Code Case N-729 from N-729-1 to N-729-4 for the 
reasons previously set forth. Due to the incorporation of N-729-4, the 
date to establish applicability for licensed pressurized water reactors 
will be changed to the effective date of the final rule.

10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6)

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(2) through (6) 
to remove the conditions currently in Sec.  50.55a(g)(6)(ii)(D)(2) 
through (5) and to redesignate the condition currently in Sec.  
50.55a(g)(6)(ii)(D)(6) as Sec.  50.55a(g)(6)(ii)(D)(2). The conditions 
currently in Sec.  50.55a(g)(6)(ii)(D)(2) to Sec.  
50.55a(g)(6)(ii)(D)(5) have all been incorporated either verbatim or 
more conservatively in the revisions to ASME BPV Code Case N-729, up to 
version N-729-4. Therefore, there is no reason to retain these 
conditions in Sec.  50.55a. The NRC proposes to include new conditions 
in Sec.  50.55a(g)(6)(ii)(D)(3) and (4) as described in the following 
discussion.

10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency

    The NRC proposes to adopt a new condition (to be included in 
proposed Sec.  50.55a(g)(6)(ii)(D)(3)) to modify the option to extend 
bare metal visual inspections of the reactor pressure vessel upper head 
surface beyond the frequency listed in Table 1 of ASME BPV Code Case N-
729-4. Previously, upper heads aged with less than eight effective 
degradation years were considered to have a low probability of 
initiating PWSCC, the cracking mechanism of concern. This ranking of 
effective degradation years was based on a simple time at temperature 
correlation. All of the upper heads within this category, with the 
exception of new heads using Alloy 600 penetration nozzles, were 
considered to have lower susceptibility to cracking due to the upper 
heads being at or near the cold leg operating temperature of the 
reactor coolant system. Therefore, these plants were said to have 
``cold heads.'' All of the upper heads that had experienced cracking 
prior to 2006 were near the hot leg operating temperature of the 
reactor coolant system, which validated the time at temperature model.

[[Page 56836]]

    In 2006, one of the 21 ``cold head'' plants identified two 
indications within a penetration nozzle and the associated partial 
penetration weld. Then, between 2006 and 2013, five of the 21 ``cold 
head'' plants identified multiple indications within fifteen different 
penetration nozzles and the associated partial penetration welds. None 
of these indications caused leakage, and volumetric examination of the 
penetration nozzles showed no flaw in the nozzle material had grown 
through-wall; however, this increasing trend creates a reasonable 
safety concern.
    Recent operational experience has shown that the volumetric 
inspection of penetration nozzles, at the current inspection frequency, 
is adequate to identify indications in the nozzle material prior to 
leakage; however, volumetric examinations cannot be performed on the 
partial penetration welds. Therefore, given the additional cracking 
identified at cold leg temperature, the NRC staff has concerns about 
the adequacy of the partial penetration weld examinations.
    Leakage from a partial penetration weld into the annulus between 
the nozzle and head material can cause corrosion of the low alloy steel 
head. While initially limited in leak rate, due to limited surface area 
of the weld being in contact with the annulus region, corrosion of the 
vessel head material can expose more of the weld surface to the 
annulus, allowing a greater leak rate. Since an indication in the weld 
cannot be identified by a volumetric inspection, a postulated crack 
through the weld, just about to cause leakage, could exist as a plant 
performed its last volumetric and/or bare metal visual examination of 
the upper head material. This gives the crack years to breach the 
surface and leak prior to the next scheduled visual examination.
    Only a surface examination of the wetted surface of the partial 
penetration weld can reliably detect flaws in the weld. Unfortunately, 
this examination cannot size the flaws in the weld, and, if performed 
manually, requires significant radiological dose to examine all the 
partial penetration welds on the upper head. As such, the available 
techniques are only able to detect a flaw after it has caused leakage. 
These techniques are a bare metal visual examination or a volumetric 
leak path assessment performed on the frequency of the volumetric 
examination.
    Volumetric leak path examinations are only done on outages when a 
volumetric examination of the nozzle is performed. Therefore, under the 
current requirements allowed by Note 4 of ASME BPV Code Case N-729-4, 
leakage from a crack in the weld of a ``cold head'' plant could start 
and continue to grow for the 5 years between the required bare metal 
visual examinations to detect leakage through the partial penetration 
weld.
    Given the additional cracking identified at cold leg temperature of 
upper head penetration nozzles and associated welds, the NRC finds 
limited basis to continue to categorize these ``cold head'' plants as 
having a low susceptibility to crack initiation. The NRC proposes to 
increase the frequency of the bare metal visual examinations of ``cold 
heads'' to identify potential leakage as soon as reasonably possible 
because of the volumetric examination limitations. Therefore, the NRC 
proposes to condition Note 4 of ASME BPV Code Case N-729-4 to require a 
bare metal visual exam each outage in which a volumetric exam is not 
performed. The NRC also proposes to allow ``cold head'' plants to 
extend their bare metal visual inspection frequency from once each 
refueling outage, as stated in Table 1 of N-729-1, to once every 5 
years, but only if the licensee performed a wetted surface examination 
of all of the partial penetration welds during the previous volumetric 
examination. Applying the conditioned bare metal visual inspection 
frequency or a volumetric examination each outage will allow licensees 
to identify any potential leakage through the partial penetration welds 
prior to significant degradation of the low alloy steel head material, 
thereby providing reasonable assurance of the structural integrity of 
the reactor coolant pressure boundary.
    These issues, including the operational experience, the fact that 
volumetric examination is not available to interrogate the partial 
penetration welds, and potential regulatory options, were discussed 
publicly at multiple ASME Code meetings, at the annual Materials 
Programs Technical Information Exchange public meeting held at the NRC 
Headquarters in June 2013, and at the 2013 NRC Regulatory Information 
Conference.

10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria

    The NRC proposes to adopt a new condition (to be included in 
proposed Sec.  50.55a(g)(6)(ii)(D)(4)) to define surface examination 
acceptance criteria. Paragraph -3132(b) of ASME BPV Code Case N-729-4 
sets forth the acceptance criteria for surface examinations. In 
general, throughout Section XI of the ASME BPV Code, the acceptance 
criteria for surface examinations default to Section III, Paragraph NB-
5352, ``Acceptance Standards''. Typically, for rounded indications, the 
indication was only unacceptable if it was greater than \3/16\ inch in 
size. The NRC requested that the code case authors include a 
requirement that any size rounded indication causing nozzle leakage is 
unacceptable due to operating experience identifying PWSCC under 
rounded indications less than \3/16\ inch in size.
    Recently, the ASME Code Committee approved an interpretation of the 
language in Paragraph -3132(b) that implied any size rounded indication 
is acceptable unless there is relevant indication of nozzle leakage, 
even those greater than \3/16\ inch. The NRC does not agree with the 
interpretation and maintains its original stance on rounded indications 
that any size rounded indication is unacceptable if there is an 
indication of leakage. Since the adoption of ASME BPV Code Case N-729-1 
into Sec.  50.55a(g)(6)(ii)(D), all licensees have used the NRC's 
stance in implementing Paragraph -3132(b), even after the recent ASME 
Code Committee interpretation approval over NRC objection.
    Therefore, in order to ensure compliance with the previous and 
ongoing requirement, the NRC proposes to revise condition Sec.  
50.55a(g)(6)(ii)(D)(4) to include clarity within the acceptance 
criteria for surface examinations. The current edition requirements of 
NB-5352 of ASME BPV Code, Section III for the licensee's ongoing 10-
year inservice inspection interval shall be met.
ASME BPV Code Case N-770-2
    On June 21, 2011, the NRC issued a final rule including Sec.  
50.55a(g)(6)(ii)(F) requiring the implementation of ASME BPV Code Case 
N-770-1, ``Alternative Examination Requirements and Acceptance 
Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds 
Fabricated with UNS N06082 or UNS N86182 Weld Filler Material With or 
Without Application of Listed Mitigation Activities,'' with certain 
conditions.
    On June 9, 2011, the ASME approved the second revision of ASME BPV 
Code Case N-770 (N-770-2). The major changes from N-770-1 to N-770-2 
included establishing new ASME Code Case Table 1 inspection item 
classifications for optimized weld overlays and allowing alternatives 
when complete inspection coverage cannot be met. Minor changes were 
also made to address editorial issues, to correct figures, or to add 
clarity. The NRC finds that the updates and improvements in N-770-2 are 
sufficient to update Sec.  50.55a(g)(6)(ii)(F).

[[Page 56837]]

    The NRC therefore proposes to update the requirements of Sec.  
50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code 
Case N-770-2 with conditions. The NRC conditions have been modified to 
address the changes in ASME BPV Code Case N-770-2 and to ensure that 
this regulatory framework will provide adequate protection of public 
health and safety. The following sections discuss each of the NRC's 
proposed changes to the conditions on ASME BPV Code Case N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(1) to change 
the version of ASME BPV Code Case N-770 from N-770-1 to N-770-2 and to 
require its implementation (with conditions) to incorporate the updates 
and improvements contained in N-770-2. The NRC proposes that licensees 
begin using N-770-2 on the effective date of this rule.

10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(2) to provide 
clarification regarding categorization of each Alloy 82/182 butt weld, 
mitigated or not, under N-770-2. This paragraph also clarifies the 
NRC's position that paragraph -1100(e) shall not be used to exempt 
welds that rely on Alloy 82/182 for structural integrity from more 
frequent ISI schedules until the NRC has reviewed and authorized an 
alternative categorization for the weld. Additionally, the NRC proposes 
to change the inspection item categories for full structural weld 
overlays from C to C-1 and F to F-1 due to reclassification under ASME 
BPV Code Case N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(3) to clarify 
the baseline examination requirements by stating that previously-
conducted examinations, in order to count as baseline examinations, 
must meet the requirements of ASME BPV Code Case N-770-2, as 
conditioned. The 2011 rule required the use of ASME Code Section XI 
Appendix VIII qualifications for baseline examinations, which is 
stricter than N-770-2 and does not provide requirements for optimized 
weld overlays. The revision also updates the deadline for baseline 
examination requirements since the January 20, 2012, deadline from the 
previous rule has passed. Finally, upon implementation of this rule, if 
a licensee is currently in an outage, then the baseline inspection 
requirement can be met by performing the inspections in accordance with 
the current regulatory requirements of Sec.  50.55a(g)(6)(ii)(F) in 
lieu of the examination requirements of paragraphs -2500(a) or -2500(b) 
of ASME BPV Code Case N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(4) to define 
examination coverage for circumferential flaws and to prohibit the use 
of paragraph -2500(d) of ASME BPV Code Case N-770-2 which, in some 
circumstances, allows unacceptably low examination coverage. Paragraph 
-2500(d) of N-770-2 would allow the reduction of circumferential 
volumetric examination coverage with analytical evaluation. Paragraph -
2500(c) was previously prohibited from use, and it continues to be 
prohibited. The NRC proposes to establish an essentially 100 percent 
volumetric examination coverage requirement for circumferential flaws 
to provide reasonable assurance of structural integrity of all ASME 
Code Class 1 butt welds susceptible to PWSCC. Therefore, the NRC 
proposes to adopt a condition prohibiting the use of paragraphs -
2500(c) and -2500(d). A licensee may request approval for use of these 
paragraphs under 10 CFR 50.55a(z).

10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(5) to add the 
explanatory heading, ``Inlay/onlay inspection frequency,'' and to make 
minor editorial corrections.

10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(6) to add the 
explanatory heading, ``Reporting requirements.''

10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(7) to add the 
explanatory heading, ``Defining `t'.''

10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(8) to maintain 
the requirement for the timing of the initial inservice examination of 
optimized weld overlays. Uncracked welds mitigated with optimized weld 
overlays were re-categorized by ASME BPV Code Case N-770-2 from 
Inspection Item D to Inspection Item C-2; however, the initial 
inspection requirement was not incorporated into the Code Case for 
Inspection Item C-2.
    The NRC has determined that uncracked welds mitigated with an 
optimized weld overlay must have an initial inservice examination no 
sooner than the third refueling outage and no later than 10 years 
following the application of the weld overlay to identify unacceptable 
crack growth. Optimized weld overlays establish compressive stress on 
the inner half thickness of the weld, but the outer half thickness may 
also be under tensile stresses. The requirement for an initial 
inservice examination no sooner than the third refueling outage and no 
later than 10 years following the application of the weld overlay is 
based on the design of optimized weld overlays which require the outer 
quarter thickness of the susceptible material to provide structural 
integrity for the weld. Therefore, the NRC proposes to continue 
adoption of the condition which requires the initial inservice 
examination of uncracked welds mitigated by optimized weld overlay 
(i.e., the welds which are subject to Inspection Item C-2 of ASME BPV 
Code Case N-770-2) within the specified timeframe.

10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(9) to address 
changes in ASME BPV Code Case N-770-2 which allow the deferral of the 
first inservice examination of uncracked welds mitigated with optimized 
weld overlays, Inspection Item C-2. Previously, under N-770-1, the 
initial inservice examination of these welds was not allowed to be 
deferred. Allowing deferral of the initial inservice examination in 
accordance with N-770-2 could, in certain circumstances, allow the 
initial inservice examination to be performed up to 20 years after 
installation. Therefore, the NRC proposes to adopt a condition which 
would preclude the deferral of the initial inservice examination of 
uncracked welds mitigated by optimized weld overlays.

10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(10) to address 
changes in ASME BPV Code Case N-770-2. Note 14(a) of Table 1 of ASME 
BPV Code Case N-770-2 provides the previously required full examination 
requirement for optimized weld overlays. The language of ASME BPV Code 
Case N-770-2, however, does not

[[Page 56838]]

require the implementation of the full examination requirements of Note 
14(a) of Table 1, if possible, before implementing the reduced 
examination coverage requirements of Note 14(b) of Table 1 or Note (b) 
of Figure 5(a). The full examination requirement should be implemented, 
if possible, before the option of reduced examination coverage is 
allowed. Therefore, the NRC proposes to modify the current condition in 
Sec.  50.55a(g)(6)(ii)(F)(10) to allow the use of Note 14(b) of Table 1 
and Note (b) of Figure 5(a) of ASME BPV Code Case N-770-2 only after 
the determination that the requirements of Note 14(a) of Table 1 of 
ASME BPV Code Case N-770-2 cannot be met.

10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(11) to address 
examination requirements through cast stainless steel materials by 
requiring the use of Appendix VIII qualifications to meet the 
inspection requirements of paragraph -2500(a) of ASME BPV Code Case N-
770-2. The requirements for volumetric examination of butt welds 
through cast stainless steel materials are currently being developed as 
Supplement 9 to the ASME BPV Code, Section XI, Appendix VIII. In 
accordance with Appendix VIII for supplements that have not been 
developed, the requirements of Appendix III apply. Appendix III 
requirements are not equivalent to Appendix VIII requirements. For the 
volumetric examination of ASME Class 1 welds, the NRC has established 
the requirement for examination qualification under the Appendix VIII. 
Thus, the NRC proposes to adopt a condition requiring the use of 
Appendix VIII qualifications to meet the inspection requirements of 
paragraph -2500(a) of ASME BPV Code Case N-770-2 by January 1, 2020.
    The development of a sufficient number of mockups would be required 
to establish an Appendix VIII program for examination of ASME Code 
Class 1 piping and vessel nozzle butt welds through cast stainless 
steel materials. The NRC recognizes that significant time and resources 
are required to create mockups and to allow for qualification of 
equipment, procedures and personnel. Therefore, the NRC proposes that 
licensees be required to use these Appendix VIII qualifications no 
later than their first scheduled weld examinations involving cast 
stainless steel materials occurring after January 1, 2020.

10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(12) to clarify 
the examination coverage requirements allowed under Appendix I of ASME 
BPV Code Case N-770-2 for butt welds joining cast stainless steel 
material. Under current ASME BPV Code, Section XI, Appendix VIII 
requirements, the volumetric examination of butt welds through cast 
stainless steel materials is under Supplement 9. Supplement 9 rules are 
still being developed by the ASME BPV Code. Therefore, it is currently 
impossible to meet the requirement of Paragraph I.5.1 for butt welds 
joining cast stainless steel material.
    The material of concern is the weld material susceptible to PWSCC 
adjoining the cast stainless steel material. Appendix VIII qualified 
procedures are available to perform the inspection of the susceptible 
weld material, but they are not qualified to inspect the cast stainless 
steel materials. Therefore, the NRC proposes to adopt a condition 
changing the inspection volume for stress-improved dissimilar metal 
welds with cast stainless steel from the ASME Code Section XI 
requirements to ``the maximum extent practical including 100 percent of 
the susceptible material volume.'' This will remain applicable until an 
Appendix VIII qualified procedure for the inspection through cast 
stainless steel materials is available in accordance with the proposed 
condition in Sec.  50.55a(g)(6)(ii)(F)(11).

10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(13) to require 
the encoding of ultrasonic volumetric examinations of Inspection Items 
A-1, A-2, B, E, F-2, J, and K in Table 1 of N-770-2. A human 
performance gap has been found between some ultrasonic testing 
procedures as demonstrated during ASME BPV Code, Section XI, Appendix 
VIII qualification versus as applied in the field.
    The human factors that contributed to the recent examinations that 
failed to identify significant flaws at North Anna Power Station, Unit 
1, in 2012 (Licensee Event Report 50-338/2012-001-00, ADAMS Accession 
No. ML12151A441) and at Diablo Canyon Nuclear Power Plant in 2013 
(Relief Request REP-1 U2, Revision 2, ADAMS Accession No. ML13232A308) 
can be avoided by the use of encoded ultrasonic examinations. Encoded 
ultrasonic examinations electronically store both the positional and 
ultrasonic information from the inspections. Encoded examinations allow 
for the inspector to evaluate the data and search for indications 
outside of a time limiting environment to assure that the inspection 
was conducted properly and to allow for sufficient time to analyze the 
data. Additionally, the encoded examination would allow for an 
independent review of the data by other inspectors or an independent 
third party. Finally, the encoded examination could be compared to 
previous and/or future encoded examinations to determine if flaws are 
present and flaw growth rate. Therefore, the NRC proposes to adopt a 
condition requiring the use of encoding for ultrasonic volumetric 
examinations of non-mitigated or cracked mitigated dissimilar metal 
butt welds in the reactor coolant pressure boundary which are within 
the scope of ASME BPV Code Case N-770-2.
ASME BPV Code Case N-824

10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii) to allow 
licensees to use the provisions of ASME BPV Code Case N-824, 
``Ultrasonic Examination of Cast Austenitic Piping Welds From the 
Outside Surface Section XI, Division 1,'' subject to NRC-proposed 
conditions of Sec.  50.55a(b)(2)(xxxvii)(A) through (E), when 
implementing inservice examinations in accordance with the ASME BPV 
Code, Section XI requirements.
    During the construction of nuclear power plants, it was recognized 
that the grain structure of cast austenitic stainless steel (CASS) 
could prevent effective ultrasonic inspections of piping welds where 
one or both sides of the welds were constructed of CASS. The high 
strength and toughness of CASS (prior to thermal embrittlement) made it 
desirable as a building material despite this known inspection issue. 
This choice of construction materials has rendered many pressure 
boundary components without a means to reliably inspect them 
volumetrically. While there is no operational experience of a CASS 
component failing, as part of the reactor pressure boundary, inservice 
volumetric inspection of these components is necessary to provide 
reasonable assurance of their structural integrity.
    The current regulatory requirements for the examination of CASS, 
provided in Sec.  50.55a, do not provide sufficient guidance to assure 
that the CASS components are being inspected

[[Page 56839]]

adequately. To illustrate that ASME Code does not provide adequate 
guidance, ASME Code, Section XI, Appendix III, Supplement 1 states 
``Cast materials may preclude meaningful examinations because of 
geometry and attenuation variables.'' For this reason, over the past 
several decades, licensees have been unable to perform effective 
inspections of welds joining CASS components. To allow for continued 
operation of their plants, licensees submitted hundreds of requests for 
relief from the ASME Code requirements for inservice inspection of CASS 
components to the NRC, resulting in a significant regulatory burden. 
Based on the improvements in ultrasonic inspection technology and 
techniques for CASS components, the ASME approved BPV Code Case N-824 
(N-824) on October 16, 2012, which describes how to develop a procedure 
capable of meaningfully inspecting welds in CASS components.
    The NRC commissioned a research program to determine the 
effectiveness of the new technologies for inspections of CASS 
components in an effort to resolve some of the known inspection issues. 
The result of this work is published in NUREG/CR-6933, ``Assessment of 
Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using 
Advanced Low-Frequency Ultrasonic Methods'', March 2007, and NUREG/CR-
7122, ``An Evaluation of Ultrasonic Phased Array Testing for Cast 
Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,'' March 
2012. These NUREG/CR reports show that CASS materials less than 1.6 
inches thick can be reliably inspected for flaws 10 percent through-
wall or deeper if encoded phased-array examinations are performed using 
low ultrasonic frequencies and a sufficient number of inspection 
angles. Additionally, for thicker welds, flaws greater than 30 percent 
through-wall in depth can be detected using low frequency encoded 
phased-array ultrasonic inspections.
    The NRC, using NUREG/CR-6933 and NUREG/CR-7122, has determined that 
inspections of CASS materials are very challenging, and sufficient 
technical basis exists to condition the code case to bring the code 
case into agreement with the NUREG/CR reports. The NUREG/CR reports 
also show that CASS materials produce high levels of coherent noise. 
The noise signals can be confusing and mask flaw indications. Use of 
encoded inspection data allows the inspector to mitigate this problem 
through the ability to electronically manipulate the data, which allows 
for discrimination between coherent noise and flaw indications. The NRC 
finds that encoding CASS inspection data provides significant detection 
benefits. The NRC proposes to add a condition in Sec.  
50.55a(b)(2)(xxxvii)(A) to require the use of encoded data when 
utilizing N-824 for the examination of CASS components. The use of dual 
element phased-array search units showed the most promise in obtaining 
meaningful responses from flaws. The NRC proposes to add a condition in 
Sec.  50.55a(b)(2)(xxxvii)(B) to require the use of dual, transmit-
receive, refracted longitudinal wave, multi-element phased array search 
units when utilizing N-824 for the examination of CASS components. The 
optimum inspection frequencies for examining CASS components of various 
thicknesses as described in NUREG/CR-6933 and NUREG/CR-7122 are 
reflected in proposed conditions Sec.  50.55a(b)(2)(xxxvii)(C) and (D). 
The NRC proposes to add a condition in Sec.  50.55a(b)(2)(xxxvii)(C) to 
require that ultrasonic examinations performed to implement ASME BPV 
Code Case N-824 on piping less than or equal to 1.6 inches thick shall 
use a phased array search unit with a center frequency of 500 kHz to 1 
MHz. The NRC proposes to add a condition in Sec.  
50.55a(b)(2)(xxxvii)(D) to require that ultrasonic examinations 
performed to implement ASME BPV Code Case N-824 on piping greater than 
1.6 inches thick shall use a phased array search unit with a center 
frequency of 500 kHz. As NUREG/CR-6933 shows that the grain structure 
of CASS can reduce the effectiveness of some inspection angles, the NRC 
finds sufficient technical basis to condition the code case for the use 
of phased-array ultrasound using angles from 30 to 70 degrees with a 
maximum increment of 5 degrees. The NRC proposes to add a condition in 
Sec.  50.55a(b)(2)(xxxvii)(E) to require that ultrasonic examinations 
performed to implement ASME BPV Code Case N-824 shall use a phased 
array search unit which produces angles from 30 to 70 degrees with a 
maximum increment of 5 degrees.
    Obtaining effective examination results of CASS components requires 
using lower frequencies and larger transducers than are typically used 
for ultrasonic inspections of piping welds and would require licensees 
to modify their inspection procedures. The NRC recognizes that 
requiring the use of spatial encoding will limit the full 
implementation of ASME BPV Code Case N-824, as spatial encoding is not 
practical for many weld configurations.
    The recent advances in inspection technology are driving renewed 
work at ASME Code meetings to produce Section XI, Appendix VIII, 
Supplement 9 to resolve the CASS inspection issue, but it will be years 
before these code updates will be published, as well as additional time 
to qualify and approve procedures for use in the field. Until then, 
licensees would still use the requirements of ASME Code Section XI, 
Appendix III, Supplement 1 which states that inspection of CASS 
materials meeting the ASME Code requirements may not be meaningful. 
Consequently, less effective examinations would continue to be used in 
the field, and more relief requests would be generated between now and 
the implementation of Supplement 9.
    At this time, the use of ASME BPV Code Case N-824, as conditioned, 
is the most effective known method for adequately examining welds with 
one or more CASS components. With the use of ASME BPV Code Case N-824, 
as conditioned, licensees will be able to take full credit for 
completion of the Sec.  50.55a required inservice volumetric inspection 
of welds involving CASS components. The implementation of ASME BPV Code 
Case N-824, as conditioned, will have the dual effect of improving the 
rigor of required volumetric inspections and reducing the number of 
uninspectable Class 1 and Class 2 pressure retaining welds.
    The NRC concludes that incorporation of ASME BPV Code Case N-824, 
as conditioned by Sec.  50.55a(b)(2)(xxxvii)(A) through (E), will 
significantly improve the flaw detection capability of ultrasonic 
inspection of CASS components until Supplement 9 is implemented, 
thereby providing reasonable assurance of leak tightness and structural 
integrity. Additionally, it will reduce the regulatory burden on 
licensees and allow licensees to submit fewer relief requests for welds 
in CASS materials.
ASME OM Code Case OMN-20

10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20

    The NRC proposes to add new paragraph Sec.  50.55a(b)(3)(x) to 
allow the use of ASME OM Code Case OMN-20, ``Inservice Test 
Frequency,'' which provides inservice test frequencies for pumps and 
valves which a licensee may voluntarily use in place of the frequencies 
specified in the 2012 Edition of the ASME OM Code. Paragraph Sec.  
50.55a(a)(1)(iii)(E) would be added to incorporate ASME OM Code Case 
OMN-20 by reference into Sec.  50.55a. Surveillance Requirement (SR) 
3.0.3 from Technical Specification (TS) 5.5.6, ``Inservice Testing 
Program,''

[[Page 56840]]

allows licensees to apply a delay period before declaring the SR for TS 
equipment ``not met'' when the licensee inadvertently exceeds or misses 
the time limit for performing TS surveillance. Licensees have been 
applying SR 3.0.3 to inservice tests. The NRC has determined that 
licensees cannot use TS 5.5.6 to apply SR 3.0.3 to inservice tests 
under Sec.  50.55a(f) that are not associated with a TS surveillance. 
To invoke SR 3.0.3, the licensee shall first discover that a TS 
surveillance was not performed at its specified frequency. Therefore, 
the delay period that SR 3.0.3 provides does not apply to non-TS 
support components tested under Sec.  50.55a(f). The ASME OM Code does 
not provide for any inservice test frequency reductions or extensions. 
In order to provide inservice test frequency reductions or extensions 
that can no longer be provided by SR 3.0.3 from TS 5.5.6, the ASME has 
developed OM Code Case OMN-20. The NRC has reviewed OM Code Case OMN-20 
and has found it acceptable for use. The NRC intends to include OM Code 
Case OMN-20 in the next revision of RG 1.192, at which time a 
conforming change will be made to delete both this paragraph and Sec.  
50.55a(a)(1)(iii)(E).

IV. Section-by-Section Analysis

    The NRC proposes to remove the revision number of the three RGs 
currently approved by the Office of the Federal Register for 
incorporation by reference throughout the substantive provisions of 
Sec.  50.55a. The revision numbers for the RGs approved for 
incorporation by reference would be retained in paragraph (a) of Sec.  
50.55a, where the RGs are listed by full title, including revision 
number. That paragraph identifies the specific materials which the 
Office of the Federal Register has approved for incorporation by 
reference, as required by Office of the Federal Register requirements 
in 1 CFR 51.9. No substantive change is intended by the NRC by this 
proposed amendment. Readers would need to refer to paragraph (a) of 
Sec.  50.55a to determine the specific revision of the relevant RG 
which is approved for incorporation by reference by Office of the 
Federal Register.

10 CFR 50.55a(a) Documents Approved for Incorporation by Reference

    The NRC proposes to revise the incorporation by reference language 
to update the contact information for the NRC Technical Library.

10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section 
III

    The NRC proposes to revise Sec.  50.55a(a)(1)(i) to clarify that 
Section III Nonmandatory Appendices are not incorporated by reference. 
This language was originally added in a final rule published on June 
21, 2011 (76 FR 36232); however, it was omitted from the final rule 
published on November 5, 2014 (79 FR 65776). The NRC is correcting the 
omission by inserting ``(excluding Nonmandatory Appendices)'' in 10 CFR 
50.55a(a)(1)(i).

10 CFR 50.55a(a)(1)(i)(E) ``Rules for Construction of Nuclear Facility 
Components--Division 1''

    The NRC proposes to revise Sec.  50.55a(a)(1)(i)(E) to add ASME BPV 
Code, Section III 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 
Edition.

10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section 
XI

    The NRC proposes to revise Sec.  50.55a(a)(1)(ii) to include a 
minor editorial change and to clarify that Nonmandatory Appendix U is 
not incorporated by reference.

10 CFR 50.55a(a)(1)(ii)(C) ``Rules for Inservice Inspection of Nuclear 
Power Plant Components--Division 1''

    The NRC proposes to revise Sec.  50.55a(a)(1)(ii)(C) to add ASME 
BPV Code, Section XI 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 
Edition.

10 CFR 50.55a(a)(1)(iii)(B) ASME BPV Code Case N-729-4

    The NRC proposes to revise Sec.  50.55a(a)(1)(iii)(B) to add the 
title ``ASME BPV Code Case N-729-4,'' and include information for the 
standard that is being incorporated by reference.

10 CFR 50.55a(a)(1)(iii)(C) ASME BPV Code Case N-770-2

    The NRC proposes to revise Sec.  50.55a(a)(1)(iii)(C) to add the 
title ``ASME BPV Code Case N-770-2,'' and include information for the 
standard that is being incorporated by reference.

10 CFR 50.55a(a)(1)(iii)(D) ASME BPV Code Case N-824

    The NRC proposes to add Sec.  50.55a(a)(1)(iii)(D) to add the title 
``ASME BPV Code Case N-824,'' and include information for the standard 
that is being incorporated by reference.

10 CFR 50.55a(a)(1)(iii)(E) ASME OM Code Case OMN-20

    The NRC proposes to add Sec.  50.55a(a)(1)(iii)(E) to add the title 
``ASME OM Code Case OMN-20,'' and include information for the standard 
that is being incorporated by reference.

10 CFR 50.55a(a)(1)(iv) ASME Operation and Maintenance Code

    The NRC proposes to revise Sec.  50.55a(a)(1)(iv) to correct the 
title of the OM Code.

10 CFR 50.55a(a)(1)(iv)(B) ``Operation and Maintenance of Nuclear Power 
Plants, Division 1: Section IST Rules for Inservice Testing of Light-
Water Reactor Power Plants''

    The NRC proposes to revise Sec.  50.55a(a)(1)(iv)(B) to add ASME OM 
Code 2009 Edition and 2011 Addenda.

10 CFR 50.55a(a)(1)(iv)(C) ``Operation and Maintenance of Nuclear Power 
Plants, Division 1: OM Code: Section IST''

    The NRC proposes to add Sec.  50.55a(a)(1)(iv)(C) to add ASME OM 
Code 2012 Edition.

10 CFR 50.55a(a)(1)(v) ASME Quality Assurance Requirements

    The NRC proposes to add Sec.  50.55a(a)(1)(v) to add the title 
``ASME Quality Assurance Requirements'' for ASME NQA-1 Code as part of 
NRC titling convention and include information regarding NQA-1 
standards.

10 CFR 50.55a(b) Use and Conditions on the Use of Standards

    The NRC proposes to revise Sec.  50.55a(b) to correct the title of 
the OM Code.

10 CFR 50.55a(b)(1) Conditions on ASME BPV Code Section III

    The NRC proposes to revise Sec.  50.55a(b)(1) to reflect the latest 
edition incorporated by reference, the 2013 Edition.

10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions

    The NRC proposes to revise Sec.  50.55a(b)(1)(ii) to clarify rule 
language and add Table 1, which clarifies prohibited Section III 
provisions in tabular form for welds with leg size less than 1.09 
tn.

10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance

    The NRC proposes to revise Sec.  50.55a(b)(1)(iv) to clarify that 
it allows, but does not require, applicants and licensees to use the 
2008 Edition through the 2009-1a Addenda of NQA-1 when applying the 
2010 Edition and later editions of the ASME BPV Code, Section III, up 
to the 2011 Addenda.

[[Page 56841]]

Applicants and licensees are required to meet appendix B of 10 CFR part 
50, and NQA-1 is one way of meeting portions of appendix B. An 
applicant or licensee may select any version of NQA-1 that has been 
approved for use in Sec.  50.55a, but they must also use the 
administrative, quality, and technical provisions contained in the 
version of NCA-4000 referencing that Edition or Addenda of NQA-1 
selected by the applicant or licensee.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1, as modified and supplemented 
by NCA-4000, does not meet all of the requirements of appendix B to 10 
CFR part 50. To meet the requirements of appendix B, when using NQA-1 
during the design and construction phase, applicants and licensees must 
address in their quality program description those areas where NQA-1 is 
insufficient to meet appendix B. Regulatory Guide 1.28, ``Quality 
Assurance Criteria (Design and Construction),'' provides additional 
guidance and regulatory positions on how to meet appendix B when using 
NQA-1.
    Section 50.55a(b)(1)(iv) clarifies that applicants and licensees 
are required to meet appendix B to 10 CFR part 50 and that the 
commitments contained in their QA program descriptions that are more 
stringent than those contained in NQA-1 or are not addressed in NQA-1 
apply to Section III activities.

10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification 
and Demonstration of Function of Incompressible-Fluid Pressure-Relief 
Valves

    The NRC proposes to revise Sec.  50.55a(b)(1)(vii) to reflect the 
latest edition incorporated by reference, the 2013 Edition.

10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME 
Certification Marks

    The NRC proposes to add Sec.  50.55a(b)(1)(viii) to allow licensees 
to use either the ASME BPV Code Symbol Stamp or ASME Certification Mark 
with the appropriate certification designator and class designator as 
specified in the 2013 Edition through the latest edition and addenda 
incorporated by reference in 10 CFR 50.55a.

10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI

    The NRC proposes to revise Sec.  50.55a(b)(2) to reflect the latest 
edition incorporated by reference, the 2013 Edition, and to clarify 
that Nonmandatory Appendix U is not incorporated by reference.

10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and 
Addenda of Subsection IWE and Subsection IWL

    The NRC proposes to revise Sec.  50.55a(b)(2)(vi) to clarify that 
the provision applies only to the class of licensees of operating 
reactors that were required by previous versions of Sec.  50.55a to 
develop and implement a containment inservice inspection program in 
accordance with Subsection IWE and Subsection IWL, and complete an 
expedited examination of containment during the 5-year period from 
September 9, 1996 to September 9, 2001.

10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment 
Examinations

    The NRC proposes to revise Sec.  50.55a(b)(2)(viii) by removing the 
condition for using the 2009 Addenda up to and including the 2013 
Edition of Subsection IWL requiring compliance with Sec.  
50.55a(b)(2)(viii)(E).

10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth 
Provision

    The NRC proposes to add Sec.  50.55a(b)(2)(viii)(H) to require 
licensees to provide the applicable information specified in paragraphs 
(b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and (b)(2)(viii)(E)(3) of this 
section in the ISI Summary Report required by IWA-6000 for each 
inaccessible concrete surface area evaluated under the new code 
provision IWL-2512 of the 2009 Addenda up to and including the 2013 
Edition.

10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth 
Provision

    The NRC proposes to add Sec.  50.55a(b)(2)(viii)(I) containing a 
new condition requiring the technical evaluation required by IWL-
2512(b) of the 2009 Addenda up to and including the 2013 Edition of 
inaccessible below-grade concrete surfaces exposed to foundation soil, 
backfill, or groundwater be performed at periodic intervals not to 
exceed 5 years. In addition, the licensee must examine representative 
samples of the exposed portions of the below-grade concrete, when such 
below-grade concrete is excavated for any reason. The proposed 
condition would apply only to holders of renewed licenses under 10 CFR 
part 54 during the period of extended operation (i.e., beyond the 
expiration date of the original 40-year license) of a renewed license 
when using IWL-2512(b) of the 2007 Edition with 2009 Addenda through 
the 2013 Edition.

10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment 
Examinations

    The NRC proposes to revise Sec.  50.55a(b)(2)(ix) to continue to 
apply the existing conditions in Sec.  50.55a(b)(2)(ix)(A)(2), Sec.  
50.55a(b)(2)(ix)(B) and Sec.  50.55a(b)(2)(ix)(J) with respect to the 
metal containment examination requirements in Subsection IWE to the 
2009 Addenda up to and including the 2013 Edition and to make minor 
editorial corrections.

10 CFR 50.55a(b)(2)(ix)(D) Metal Containment Examinations: Fourth 
Provision

    The NRC proposes to revise the rule text in Sec.  
50.55a(b)(2)(ix)(D) to improve clarity. Paragraphs Sec.  
50.55a(b)(2)(ix)(D) and Sec.  50.55a(b)(2)(ix)(D)(1) are combined. The 
information required to be included in the ISI Summary report is now 
all on the same paragraph level. No substantive change to the 
requirements is intended by this revision.

10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance

    The NRC proposes to revise Sec.  50.55a(b)(2)(x) to clarify that it 
allows, but does not require, licensees to use the 1994 or the 2008 
Edition through the 2009-1a Addenda of NQA-1 when applying the 2009 
Addenda and later editions and addenda of the ASME BPV Code, Section 
XI, up to the 2013 Edition. Licensees are required to meet appendix B 
of 10 CFR part 50, and NQA-1 is one way of meeting portions of appendix 
B. A licensee may select any version of NQA-1 that has been approved 
for use in Sec.  50.55a.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1 does not meet all of the 
requirements of appendix B to 10 CFR part 50. To meet the requirements 
of appendix B, when using NQA-1 during inservice inspection phase, 
licensees must address in their quality program description those areas 
where NQA-1 is insufficient to meet appendix B. Additional guidance and 
regulatory positions on how to meet appendix B when using NQA-1 is 
provided in RG 1.28, ``Quality Assurance Criteria (Design and 
Construction).''

[[Page 56842]]

    Section 50.55a(b)(2)(x) clarifies that licensees are required to 
meet appendix B to 10 CFR part 50 and that the commitments contained in 
their QA program descriptions that are more stringent than those 
contained in NQA-1 or are not addressed in NQA-1 apply to Section XI 
activities.

10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth 
Provision

    The NRC proposes to add Sec.  50.55a(b)(2)(xviii)(D) to provide a 
new condition prohibiting the use of Appendix VII and subarticle VIII-
2200 of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV 
Code. Licensees would be required to implement Appendix VII and 
subarticle VIII-2200 of the 2010 Edition of Section XI.

10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements: 
First Provision

    The NRC proposes to revise Sec.  50.55a(b)(2)(xxi)(A) to modify the 
standard for visual magnification resolution sensitivity and contrast 
for visual examinations performed on Examination Category B-D 
components instead of ultrasonic examinations. A visual examination 
with magnification that has a resolution sensitivity to resolve 0.044 
inch (1.1 mm) lower case characters without an ascender or descender 
(e.g., a, e, n, v), utilizing the allowable flaw length criteria in 
Table IWB-3512-1, 1997 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, with 
a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may 
be performed instead of an ultrasonic examination. This revision 
removes a requirement that was in addition to ASME BPV Code that 
required 1-mil wires to be used in licensees' Sensitivity, Resolution 
and Contrast Standard targets.

10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator 
Preservice Examinations

    The NRC proposes to add Sec.  50.55a(b)(2)(xxx) to provide a new 
condition requiring that instead of the preservice inspection 
requirements of Section XI, IWB-2200(c), a full length examination of 
100 percent of the tubing in each newly installed steam generator shall 
be performed prior to plant startup. These inspections shall be 
performed with the objective of finding the types of flaws that may 
potentially be present in the tubes and that may potentially occur 
during operation.

10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping 
Devices

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxi) to provide a new 
condition prohibiting the use of mechanical clamping devices in 
accordance with IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d) in 
the 2011 Addenda through 2013 Edition on small item Class 1 piping and 
portions of a piping system that forms the containment boundary.

10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report 
Submittal

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxii) to provide a new 
condition requiring licensees using the 2010 Edition or later editions 
and addenda of Section XI to follow the requirements of IWA-6240 of the 
2009 addenda of Section XI for the submittal of Preservice and 
Inservice Summary Reports.

10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed 
Allowable Pressure

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxiii) to provide a new 
condition to prohibit the use of Appendix G Paragraph G-2216 in the 
2011 Addenda and later editions and addenda of the ASME BPV Code, 
Section XI.

10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Disposition of Flaws 
in Class 3 Components

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxiv) to provide a new 
condition to require that when using the 2013 Edition of the ASME BPV 
Code, Section XI, the licensee shall use the acceptance standards of 
IWD-3510 for the disposition of flaws in Category D-A components (i.e., 
welded attachments for vessels, piping, pumps, and valves).

10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 
in the KIa and KIc Equations

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxv) to provide a new 
condition to specify that when licensees use ASME BPV Code, Section XI, 
2013 Edition Appendix A paragraph A-4200, if T0 is 
available, then RTT0 may be used in place of 
RTNDT for applications using the KIc equation and 
the associated KIc curve, but not for applications using the 
KIa equation and the associated KIa curve.

10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of 
Irradiated Materials

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvi) to provide a new 
condition requiring licensees using ASME BPV Code, Section XI, 2013 
Edition, Appendix A, paragraph A-4400, to obtain NRC approval before 
using irradiated T0 and the associated RTT0 in 
establishing fracture toughness of irradiated materials.

10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii) with 
subparagraphs (A) through (E) to provide a new provision that allows 
licensees to implement ASME BPV Code Case N-824, ``Ultrasonic 
Examination of Cast Austenitic Piping Welds From the Outside Surface 
Section XI, Division 1,'' as conditioned by subparagraphs (A) through 
(E).

10 CFR 50.55a(b)(2)(xxxvii)(A) Section XI Condition: ASME BPV Code Case 
N-824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii)(A) to add a new 
condition that requires ultrasonic examinations performed to implement 
ASME BPV Code Case N-824 to be spatially encoded.

10 CFR 50.55a(b)(2)(xxxvii)(B) Section XI Condition: ASME BPV Code Case 
N-824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii)(B) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 shall use dual, transmit-receive, 
refracted longitudinal wave, multi-element phased array search units 
instead of the requirements of Paragraph 1(c)(1)(-a) of N-824.

10 CFR 50.55a(b)(2)(xxxvii)(C) Section XI Condition: ASME BPV Code Case 
N-824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii)(C) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 on piping less than or equal to 1.6 
inches thick shall use a phased array search unit with a center 
frequency of 500 kHz to 1 MHz instead of the requirements of Paragraph 
1(c)(1)(-c)(-1).

10 CFR 50.55a(b)(2)(xxxvii)(D) Section XI Condition: ASME BPV Code Case 
N-824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii)(D) to add a new

[[Page 56843]]

condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 on piping greater than 1.6 inches 
thick shall use a phased array search unit with a center frequency of 
500 kHz instead of the requirements of Paragraph 1(c)(1)(-c)(-2).

10 CFR 50.55a(b)(2)(xxxvii)(E) Section XI Condition: ASME BPV Code Case 
N-824

    The NRC proposes to add Sec.  50.55a(b)(2)(xxxvii)(E) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 shall use a phased array search unit 
which produces angles from 30 to 70 degrees with a maximum increment of 
5 degrees instead of the requirements of Paragraph 1(c)(1)(-d).

10 CFR 50.55a(b)(3) Conditions on ASME OM Code

    The NRC proposes to revise Sec.  50.55a(b)(3) to require that the 
2012 Edition of the ASME OM Code be used during the initial 120-month 
inservice test interval under Sec.  50.55a(f)(4)(i) and during 
mandatory 120-month IST program updates under Sec.  50.55a(f)(4)(ii). 
The proposed revision would also allow users to voluntarily update 
their IST programs to the 2009 Edition, 2011 Addenda, or 2012 Edition 
of the ASME OM Code (with the exceptions and conditions specified in 
this notice) under Sec.  50.55a(f)(4)(iv).

10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance

    The NRC proposes to revise Sec.  50.55a(b)(3)(i) to allow licensees 
to use the 1983 Edition through the 1994 Edition, 2008 Edition, and 
2009-1a Addenda of NQA-1 when using the 1995 Edition through the 2012 
Edition of the ASME OM Code. Licensees are required to meet appendix B 
to 10 CFR part 50, and NQA-1 is one way of meeting portions of appendix 
B.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1 does not meet all of the 
requirements of appendix B to 10 CFR part 50. To meet the requirements 
of appendix B, licensees must address in their quality program 
description those areas where NQA-1 is insufficient to meet appendix B. 
Regulatory Guide 1.28, ``Quality Assurance Criteria (Design and 
Construction),'' provides additional guidance on how to meet appendix B 
when using NQA-1.
    Paragraph 50.55a(b)(3)(i) clarifies that licensees are required to 
meet appendix B to 10 CFR part 50 and that the commitments contained in 
their QA program descriptions that are more stringent than those 
contained in NQA-1 or are not addressed in NQA-1 apply to OM Code 
activities.

10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) 
Testing

    The NRC proposes to revise Sec.  50.55a(b)(3)(ii) to reflect 
Appendix III, ``Preservice and Inservice Testing of Active Electric 
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,'' 
in the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition.

10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval

    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(A) to require that 
licensees evaluate the adequacy of the diagnostic test interval for 
each MOV and adjust the interval as necessary, but not later than 5 
years or three refueling outages (whichever is longer) from initial 
implementation of Appendix III of the ASME OM Code.

10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk

    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(B) to require that 
licensees ensure that the potential increase in core damage frequency 
and large early release frequency associated with the extension is 
acceptably small when extending exercise test intervals for high risk 
MOVs beyond a quarterly frequency.

10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization

    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(C) to require, when 
applying Appendix III to the ASME OM Code, that licensees categorize 
MOVs according to their safety significance using the methodology 
described in ASME OM Code Case OMN-3 subject to the conditions 
discussed in RG 1.192, or using an MOV risk ranking methodology 
accepted by the NRC on a plant-specific or industry-wide basis in 
accordance with the conditions in the applicable safety evaluation.

10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time

    The NRC proposes to add Sec.  50.55a(b)(3)(ii)(D) to require, when 
applying Paragraph III-3600, ``MOV Exercising Requirements,'' of 
Appendix III to the OM Code, licensees shall verify that the stroke 
time of the MOV satisfies the assumptions in the plant safety analyses.

10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors

    The NRC proposes to add Sec.  50.55a(b)(3)(iii) to specify that, in 
addition to complying with the provisions in the OM Code as required 
with the conditions specified in Sec.  50.55a(b)(3), holders of 
operating licenses for nuclear power reactors that received 
construction permits under this part on or after the date 12 months 
after the effective date of this rulemaking and holders of COLs issued 
under 10 CFR part 52, whose initial fuel loading occurs on or after the 
date 12 months after the effective date of this rulemaking, shall also 
comply with specified conditions, as applicable.

10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves

    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(A) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) develop a program to 
periodically verify the capability of power-operated valves (POVs) to 
perform their design-basis safety functions.

10 CFR 50.55a(b)(3)(iii)(B) Check Valves

    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(B) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) perform bi-directional 
testing of check valves within the IST program where practicable.

10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration

    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(C) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) monitor flow-induced 
vibration (FIV) from hydrodynamic loads and acoustic resonance during 
preservice testing and inservice testing to identify potential adverse 
flow effects that might impact components within the scope of the IST 
program.

10 CFR 50.55a(b)(3)(iii)(D) High Risk Non-Safety Systems

    The NRC proposes to add Sec.  50.55a(b)(3)(iii)(D) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) establish a program to 
assess the operational readiness of pumps, valves, and dynamic 
restraints within the scope of the Regulatory Treatment of Non-Safety 
Systems (RTNSS) for applicable reactor designs. The proposed rule 
language refers to such components using the term, ``high risk non-
safety systems.''

[[Page 56844]]

10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)

    The NRC proposes to revise Sec.  50.55a(b)(3)(iv) to specify that 
Appendix II in the 2003 Addenda through the 2012 Edition of the OM Code 
is acceptable for use without conditions with the clarifications that 
(1) the maximum test interval allowed by Appendix II for individual 
check valves in a group of two valves or more must be supported by 
periodic testing of a sample of check valves in the group during the 
allowed interval and (2) the periodic testing plan must be designed to 
test each valve of a group at approximate equal intervals not to exceed 
the maximum requirement interval. The conditions currently specified 
for the use of Appendix II, 1995 Edition with the 1996 and 1997 
Addenda, and 1998 Edition through the 2002 Addenda, of the OM Code 
remain the same in this proposed rule.

10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB

    The NRC proposes to add Sec.  50.55a(b)(3)(vii) to prohibit the use 
of Subsection ISTB in the 2011 Addenda to the ASME OM Code.

10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE

    The NRC proposes to add Sec.  50.55a(b)(3)(viii) to specify that 
licensees who wish to implement Subsection ISTE, ``Risk-Informed 
Inservice Testing of Components in Light-Water Reactor Nuclear Power 
Plants,'' of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 
Edition, must first request and obtain NRC approval in accordance with 
Sec.  50.55a(z) to apply Subsection ISTE on a plant-specific basis as a 
risk-informed alternative to the applicable IST requirements in the 
ASME OM Code.

10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF

    The NRC proposes to add Sec.  50.55a(b)(3)(ix) to specify that 
licensees applying Subsection ISTF, ``Inservice Testing of Pumps in 
Light-Water Reactor Nuclear Power Plants--Post-2000 Plants,'' in the 
2012 Edition of the OM Code shall satisfy the requirements of Mandatory 
Appendix V, ``Pump Periodic Verification Test Program,'' of the OM 
Code, 2012 Edition. The proposed paragraph will also state that 
Subsection ISTF, 2011 Addenda, is not acceptable for use.

10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20

    The NRC proposes to add Sec.  50.55a(b)(3)(x) to allow licensees to 
implement ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' in 
the ASME OM Code, 2012 Edition.

10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication

    The NRC proposes to add Sec.  50.55a(b)(3)(xi) to require that 
licensees supplement the ASME OM Code provisions in Subsection ISTC-
3700, ``Position Verification Testing,'' as necessary to verify that 
valve operation is accurately indicated. The ASME OM Code, Subsection 
ISTC-3700 requires valves with remote position indicators shall be 
observed locally at least once every 2 years to verify that valve 
operation is accurately indicated.

10 CFR 50.55a(f): Inservice Testing Requirements

    The NRC proposes to revise Sec.  50.55a(f) to clarify that the ASME 
OM Code includes provisions for preservice testing of components as 
part of its overall provisions for IST programs.

10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision

    The NRC proposes to revise Sec.  50.55a(f)(3)(iii)(A) to state that 
the paragraph is applicable to pumps and valves that are within the 
scope of the ASME OM Code. This will align the scope of pumps and 
valves for inservice testing with the scope defined in the ASME Code 
and in SRP Section 3.9.6.

10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision

    The NRC proposes to revise Sec.  50.55a(f)(3)(iii)(B) to ensure 
that the paragraph is applicable to pumps and valves that are within 
the scope of the ASME OM Code. This will align the scope of pumps and 
valves for inservice testing with the scope defined in the ASME Code 
and in SRP Section 3.9.6.

10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First 
Provision

    The NRC proposes to revise Sec.  50.55a(f)(3)(iv)(A) to ensure that 
the paragraph is applicable to pumps and valves that are within the 
scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(A) 
for Class 1 pumps and valves. This will align the scope of pumps and 
valves for inservice testing with the scope defined in the ASME Code 
and in SRP Section 3.9.6.

10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second 
Provision

    The NRC proposes to revise Sec.  50.55a(f)(3)(iv)(B) to ensure that 
the paragraph is applicable to pumps and valves that are within the 
scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(B) 
for Class 1 pumps and valves. This will align the scope of pumps and 
valves for inservice testing with the scope defined in the ASME Code 
and in SRP Section 3.9.6.

10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for 
Operating Plants

    The NRC proposes to revise Sec.  50.55a(f)(4) to ensure that the 
paragraph is applicable to pumps and valves that are within the scope 
of the ASME OM Code. This will align the scope of pumps and valves for 
inservice testing with the scope defined in the ASME Code and in SRP 
Section 3.9.6.

10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements

    The NRC proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and 
(g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i), 
(g)(3)(ii), and (g)(3)(v) to distinguish the requirements for 
accessibility and preservice examination from those for inservice 
inspection in Sec.  50.55a(g). No substantive change to the 
requirements is intended by these revisions.

10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(1) to require 
licensees to implement an augmented inservice inspection program for 
the examination of the RPV upper head penetrations meeting ASME BPV 
Code Case N-729-4 instead of the previously approved requirements to 
use ASME BPV Code Case N-729-1, as conditioned by the NRC.

10 CFR 50.55a(g)(6)(ii)(D)(2) Through (5) of the Current Regulation

    The NRC proposes to remove the conditions in existing Sec.  
50.55a(g)(6)(ii)(D)(2) through (5) of the current regulation, inasmuch 
as these conditions have been included in or reflected in other Code 
requirements. In their place, the NRC proposes to adopt new conditions 
in Sec.  50.55a(g)(6)(ii)(D)(2) through (4).

10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I Use

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(2) to require 
NRC approval prior to implementing Appendix I of ASME BPV Code Case N-
729-4. This requirement is currently located in Sec.  
50.55a(g)(6)(ii)(D)(6) for implementation of N-729-1.

[[Page 56845]]

10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(3) to add a 
new condition which requires cold head plants (EDY<8) without PWSCC 
flaws to perform a bare metal visual examination (VE) each outage a 
volumetric exam is not performed and allows these plants to extend the 
bare metal visual inspection frequency from once each refueling outage, 
as stated in Table 1 of N-729-4, to once every 5 years only if the 
licensee performed a wetted surface examination of all of the partial 
penetration welds during the previous volumetric examination. In 
addition, this new condition clarifies that a bare metal visual 
examination is not required during refueling outages when a volumetric 
or surface examination is performed of the partial penetration welds. 
The condition that is in the current Sec.  50.55a(g)(6)(ii)(D)(3) was 
incorporated into N-729-4 by the ASME Code committees.

10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(D)(4) to add a 
new condition which clarifies that rounded indications found by surface 
examinations of the partial-penetration or associated fillet welds in 
accordance with N-729-4 must meet the acceptance criteria for surface 
examinations of paragraph NB-5352 of ASME Section III of the current 
edition and addenda for the licensee's ongoing 10-year inservice 
inspection interval. The condition that is in the current Sec.  
50.55a(g)(6)(ii)(D)(4) was incorporated into N-729-4 by the ASME Code 
committees.

10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(1) to require 
licensees to implement an augmented inservice inspection program for 
the examination of ASME Class 1 piping and nozzle butt welds meeting 
ASME BPV Code Case N-770-2 instead of the previously approved ASME BPV 
Code Case N-770-1.
    Furthermore, the NRC proposes to revise Sec.  
50.55a(g)(6)(ii)(F)(1) to update the date of applicability for 
pressurized water reactors, to note the change to implement ASME BPV 
Code Case N-770-2 instead of N-770-1, and to reflect the number of 
conditions which must be applied.

10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(2) to clarify 
the requirements for licensees to establish the initial categorization 
of each weld and modify the wording to reflect the ASME BPV Code Case 
N-770-2 change in the inspection item category for full structural weld 
overlays. Additionally, the NRC proposes to add a sentence which 
clarifies the NRC position that paragraph -1100(e) of ASME BPV Code 
Case N-770-2 shall not be used to exempt welds that rely on Alloy 82/
182 for structural integrity from any requirement of Sec.  
50.55a(g)(6)(ii)(F).

10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(3) to clarify 
the current requirement in this paragraph to complete baseline 
examinations. Additionally, this condition clarifies that the 
examination coverage requirements, for a licensee to count previous 
inspections as baseline examinations, are the same examination coverage 
requirements described in paragraphs -2500(a) or -2500(b) of ASME BPV 
Code Case N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(4) to clarify 
that licensees are required to ensure greater than 90 percent 
volumetric examination coverage is obtained for circumferential flaws, 
to continue the restriction on the licensee's use of paragraph -2500(c) 
and to continue the restriction that the use of new paragraph -2500(d) 
of ASME BPV Code Case N-770-2 is not allowed without prior NRC review 
and approval in accordance with Sec.  50.55a(z), as it would permit a 
reduction in volumetric examination coverage for circumferential flaws.

10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(5) to add 
explanatory heading and to make minor editorial corrections.

10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(6) to add 
explanatory heading.

10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(7) to add 
explanatory heading.

10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(8) to continue 
the current condition located in Sec.  50.55a(g)(6)(ii)(F)(9) which 
requires that the initial examination of optimized weld overlays (i.e., 
Inspection Item C-2 of ASME BPV Code Case N-770-2) be performed between 
the third refueling outage and no later than 10 years after application 
of the overlay and delete the other current examination requirements 
for optimized weld overlay examination frequency, as these requirements 
were included in the revision from N-770-1 to N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(9) to modify 
the current condition to continue denial of the deferral of the initial 
inservice examination of uncracked welds mitigated by optimized weld 
overlays. These welds shall continue to have their initial inservice 
examinations as prescribed in N-770-1 within 10 years of the 
application of the optimized weld overlay and not allow deferral of 
this initial examination. Subsequent inservice examinations may be 
deferred as allowed by N-770-2. Additionally, the modified condition 
will delete the current condition on examination requirements for the 
deferral of welds mitigated by inlay, onlay, stress improvement and 
optimized weld overlay, as these requirements were, with one exception 
(i.e., optimized weld overlay), included in the revision from N-770-1 
to N-770-2.

10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique

    The NRC proposes to revise Sec.  50.55a(g)(6)(ii)(F)(10) to modify 
the current condition to allow the previously prohibited alternate 
examination requirements of Note (b) of Figure 5(a) of ASME BPV Code 
Case N-770-1 and N-770-2 and the same requirements in Note 14(b) of 
Table 1 of ASME BPV Code Case N-770-2 for optimized weld overlays only 
if the full examination requirements of Note 14(a) of Table 1 of ASME 
BPV Code Case N-770-2 cannot be met.

10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(11) to provide a 
new condition requiring licensees to establish a Section XI Appendix 
VIII qualification requirement for ultrasonic inspection of and through 
cast stainless steel to meet the examination requirements of paragraph 
-2500(a) of

[[Page 56846]]

ASME BPV Code Case N-770-2 by January 1, 2020.

10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(12) to provide a 
new condition that would allow licenses to implement a stress 
improvement mitigation technique for items containing cast stainless 
steel that would meet the requirements of Appendix I of ASME BPV Code 
Case N-770-2, if the required examination volume can be examined by 
Appendix VIII procedures to the maximum extent practical including 100 
percent of the susceptible material volume.

10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination

    The NRC proposes to add Sec.  50.55a(g)(6)(ii)(F)(13) to provide a 
new condition requiring licensees to perform encoded examinations of 
essentially 100 percent of the inspection surface area when required to 
perform volumetric examinations of all non-mitigated and cracked 
mitigated butt welds in accordance with N-770-2.

V. Generic Aging Lessons Learned Report

Background

    In December 2010, the NRC issued ``Generic Aging Lessons Learned 
(GALL) Report,'' NUREG-1801, Revision 2, for applicants to use in 
preparing their license renewal applications. The GALL Report provides 
aging management programs (AMPs) that the NRC staff has concluded are 
sufficient for aging management in accordance with the license renewal 
rule, as required in 10 CFR 54.21(a)(3). In addition, ``Standard Review 
Plan for Review of License Renewal Applications for Nuclear Power 
Plants,'' NUREG-1800, Revision 2 was issued in December 2010 to ensure 
the quality and uniformity of NRC staff reviews of license renewal 
applications and to present a well-defined basis on which the NRC staff 
evaluates the applicant's aging management programs and activities. In 
April 2011, the NRC also issued ``Disposition of Public Comments and 
Technical Bases for Changes in the License Renewal Guidance Documents 
NUREG-1801 and NUREG-1800,'' NUREG-1950, which describes the technical 
bases for the changes in Revision 2 of the GALL Report and Revision 2 
of the SRP for review of license renewal applications.
    Revision 2 of the GALL Report, in Sections XI.M1, XI.S1, XI.S2, and 
XI.S3, describes the evaluation and technical bases for determining the 
sufficiency of ASME BPV Code Subsections IWB, IWC, IWD, IWE, IWF, and 
IWL for managing aging during the period of extended operation. In 
addition, many other aging management programs in the GALL Report rely, 
in part but to a lesser degree, on the requirements specified in the 
ASME BPV Code, Section XI. Revision 2 of the GALL Report also states 
that the 1995 Edition through the 2004 Edition of the ASME BPV Code, 
Section XI, Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as modified 
and limited by Sec.  50.55a, were found to be acceptable editions and 
addenda for complying with the requirements of 10 CFR 54.21(a)(3), 
unless specifically noted in certain sections of the GALL Report. The 
GALL Report further states that the future Federal Register notices 
that amend Sec.  50.55a will discuss the acceptability of editions and 
addenda more recent than the 2004 edition for their applicability to 
license renewal. In a final rule issued on June 21, 2011 (76 FR 36232), 
subsequent to Revision 2 of the GALL Report, the NRC also found that 
the 2004 Edition with the 2005 Addenda through the 2007 Edition with 
the 2008 Addenda of Section XI of the ASME BPV Code, Subsections IWB, 
IWC, IWD, IWE, IWF, and IWL, as subject to the conditions in Sec.  
50.55a, are acceptable for the AMPs in the GALL Report and the 
conclusions of the GALL Report remain valid with the augmentations 
specifically noted in the GALL Report.

Evaluation With Respect to Aging Management

    As part of this rulemaking, the NRC evaluated whether those AMPs in 
Revision 2 of the GALL Report which rely upon Subsections IWB, IWC, 
IWD, IWE, IWF, and IWL of Section XI in the editions and addenda of the 
ASME BPV Code incorporated by reference into Sec.  50.55a, continue to 
be acceptable if the AMP relies upon the versions of these Subsections 
in the 2007 Edition with the 2009 Addenda through the 2013 Edition. The 
NRC finds that the 2007 Edition with the 2009 Addenda through the 2013 
Edition of Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, 
IWE, IWF, and IWL, as subject to the conditions of this rule, are 
acceptable for the AMPs in the GALL Report and the conclusions of the 
GALL Report remain valid with the augmentations specifically noted in 
the GALL Report. Accordingly, an applicant for license renewal may use, 
in its plant-specific license renewal application, Subsections IWB, 
IWC, IWD, IWE, IWF, and IWL of Section XI of the 2007 Edition with the 
2009 Addenda through the 2013 Edition of the ASME BPV Code, as subject 
to the conditions in this rule, without additional justification. 
Similarly, a licensee approved for license renewal that relied on the 
GALL AMPs may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of 
Section XI of the 2007 Edition with the 2009 Addenda through the 2013 
Edition of the ASME BPV Code. However, a licensee must assess and 
follow applicable NRC requirements with regard to changes to its 
licensing basis.
    Some of the AMPs in the GALL Report recommend augmentation of 
certain Code requirements in order to ensure adequate aging management 
for license renewal. The technical and regulatory aspects of the AMPs 
for which augmentations are recommended also apply if the editions or 
addenda from the 2007 Edition with the 2009 Addenda through the 2013 
Edition of Section XI of the ASME BPV Code are used to meet the 
requirements of 10 CFR 54.21(a)(3). The NRC staff evaluated the changes 
in the 2007 Edition with the 2009 Addenda through the 2013 Edition of 
Section XI of the ASME BPV Code to determine if the augmentations 
described in the GALL Report remain necessary; the NRC staff's 
evaluation has concluded that the augmentations described in the GALL 
Report are necessary to ensure adequate aging management. For example, 
Table IWB-2500-1, in the 2007 Edition with the 2009 Addenda of ASME BPV 
Code, Section XI, Subsection IWB, requires surface examination of ASME 
Code Class 1 branch pipe connection welds less than nominal pipe size 
(NPS) 4 under Examination Category B-J. However, the NRC staff finds 
that volumetric or opportunistic destructive examination rather than 
surface examination is necessary to adequately detect and manage the 
aging effect due to stress corrosion cracking or thermal, mechanical 
and vibratory loadings in the components for the period of extended 
operation. Therefore, GALL Report Section XI.M35, ``One-Time Inspection 
of ASME Code Class 1 Small-Bore Piping,'' includes the augmentation of 
the requirements in ASME BPV Code, Section XI, Subsection IWB to 
perform a one-time inspection of a sample of ASME Code Class 1 piping 
less than NPS 4 and greater than or equal to NPS 1 using volumetric or 
opportunistic destructive examination. The GALL Report addresses this 
augmentation to confirm that there is no need to manage age-related 
degradation through periodic volumetric inspections or that an existing 
AMP (for example, Water

[[Page 56847]]

Chemistry AMP) is effective to manage the aging effect due to stress 
corrosion cracking or thermal, mechanical and vibratory loadings for 
the period of extended operation. A license renewal applicant may 
either augment its AMPs as described in the GALL Report, or propose 
alternatives for the NRC to review as part of the applicant's plant-
specific justification for its AMPs.

VI. Specific Request for Comments

    The NRC requests specific comments on the following questions:
    NRC Question 1. NQA-1. The NRC is considering removing the 
references to versions of NQA-1 older than the 1994 Edition in Sec.  
50.55a(b)(1)(iv), Sec.  50.55a(b)(2)(x), and Sec.  50.55a(b)(3)(i). The 
NRC requests public comment on whether any applicant or licensee is 
committed to, and is using, a version of NQA-1 older than the 1994 
Edition, and if so, what version the applicant or licensee is using.
    NRC Question 2. ASME BPV Code Case N-824. The NRC is proposing to 
make ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast 
Austenitic Piping Welds From the Outside Surface Section XI, Division 
1,'' acceptable for use with conditions. The use of N-824, as 
conditioned, is considered a stop-gap improvement until ASME Section XI 
Appendix VIII Supplement 9 is developed and implemented. The NRC is 
considering whether ASME BPV Code Case N-824, as conditioned, should be 
mandatory because of the potential that licensees may continue to use 
less effective ASME Code Section XI Appendix III techniques for 
examinations of welds next to CASS material. Should ASME BPV Code Case 
N-824, as conditioned, be mandatory? What are the possible advantages 
and disadvantages of making N-824, as conditioned, mandatory?

VII. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, and well-organized 
manner. The NRC has written this document to be consistent with the 
Plain Writing Act as well as the Presidential Memorandum, ``Plain 
Language in Government Writing,'' published June 10, 1998 (63 FR 
31883). The NRC requests comment on this document with respect to the 
clarity and effectiveness of the language used.

VIII. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113 (NTTAA), and implementing guidance in U.S. Office of 
Management and Budget (OMB) Circular A-119 (February 10, 1998), 
requires that Federal agencies use technical standards that are 
developed or adopted by voluntary consensus standards bodies unless 
using such a standard is inconsistent with applicable law or is 
otherwise impractical. The NTTAA requires Federal agencies to use 
industry consensus standards to the extent practical; it does not 
require Federal agencies to endorse a standard in its entirety. Neither 
the NTTAA nor Circular A-119 prohibit an agency from adopting a 
voluntary consensus standard while taking exception to specific 
portions of the standard, if those provisions are deemed to be 
``inconsistent with applicable law or otherwise impractical.'' 
Furthermore, taking specific exceptions furthers the Congressional 
intent of Federal reliance on voluntary consensus standards because it 
allows the adoption of substantial portions of consensus standards 
without the need to reject the standards in their entirety because of 
limited provisions that are not acceptable to the agency.
    In this rulemaking, the NRC is continuing its existing practice of 
establishing requirements for the design, construction, operation, 
inservice inspection (examination) and inservice testing of nuclear 
power plants by approving the use of the latest editions and addenda of 
the ASME BPV and OM Codes (ASME Codes) in Sec.  50.55a. The ASME Codes 
are voluntary consensus standards, developed by participants with broad 
and varied interests, in which all interested parties (including the 
NRC and licensees of nuclear power plants) participate. Therefore, the 
NRC's incorporation by reference of the ASME Codes is consistent with 
the overall objectives of the NTTAA and OMB Circular A-119.
    As discussed in Section III of this statement of considerations, in 
this proposed rule the NRC is conditioning the use of certain 
provisions of the 2009 Addenda, 2010 Edition, 2011 Addenda, and the 
2013 Edition to the ASME BPV Code, Section III, Division 1 and the ASME 
BPV Code, Section XI, Division 1, including NQA-1 (with conditions on 
its use), as well as the 2009 Edition and 2011 Addenda and 2012 Edition 
to the ASME OM Code and Code Cases N-770-2, N-729-4, and N-824. In 
addition, the proposed rule does not adopt (``excludes'') certain 
provisions of the ASME Codes and this statement of considerations, and 
in the regulatory and backfit analysis for this rulemaking. The NRC 
believes that this proposed rule complies with the NTTAA and OMB 
Circular A-119 despite these conditions and ``exclusions.''
    If the NRC did not conditionally accept ASME editions, addenda, and 
code cases, the NRC would disapprove these entirely. The effect would 
be that licensees and applicants would submit a larger number of 
requests for use of alternatives under Sec.  50.55a(z), requests for 
relief under Sec.  50.55a(f) and (g), or requests for exemptions under 
Sec.  50.12 and/or Sec.  52.7. These requests would likely include 
broad-scope requests for approval to issue the full scope of the ASME 
Code editions and addenda which would otherwise be approved as proposed 
in this rulemaking (i.e., the request would not be simply for approval 
of a specific ASME Code provision with conditions). These requests 
would be an unnecessary additional burden for both the licensee and the 
NRC, inasmuch as the NRC has already determined that the ASME Codes and 
Code Cases that are the subject of this rulemaking are acceptable for 
use (in some cases with conditions). For these reasons, the NRC 
concludes that this proposed rule's treatment of ASME Code editions and 
addenda, and code cases and any conditions placed on them does not 
conflict with any policy on agency use of consensus standards specified 
in OMB Circular A-119.
    The NRC did not identify any other voluntary consensus standards 
developed by U.S. voluntary consensus standards bodies for use within 
the U.S. that the NRC could incorporate by reference instead of the 
ASME Codes. The NRC also did not identify any voluntary consensus 
standards developed by multinational voluntary consensus standards 
bodies for use on a multinational basis that the NRC could incorporate 
by reference instead of the ASME Codes. The NRC identified codes 
addressing the same subject as the ASME Codes for use in individual 
countries. At least one country, Korea, directly translated the ASME 
Code for use in that country. In other countries (e.g., Japan), ASME 
Codes were the basis for development of the country's codes, but the 
ASME Codes were substantially modified to accommodate that country's 
regulatory system and reactor designs. Finally, there are countries 
(e.g., the Russian Federation) where that country's code was developed 
without regard to the ASME Code. However, some of these codes may not 
meet the definition of a voluntary consensus standard because they were 
developed by the state rather than a voluntary consensus standards 
body. Evaluation by the NRC of the countries' codes to determine 
whether each code provides a comparable or enhanced level of safety

[[Page 56848]]

when compared against the level of safety provided under the ASME Codes 
would require a significant expenditure of agency resources. This 
expenditure does not seem justified, given that substituting another 
country's code for the U.S. voluntary consensus standard does not 
appear to substantially further the apparent underlying objectives of 
the NTTAA.
    In summary, this proposed rulemaking satisfies the requirements of 
the NTTAA and OMB Circular A-119.

IX. Incorporation by Reference--Reasonable Availability to Interested 
Parties

    The NRC proposes to incorporate by reference seven recent editions 
and addenda to the ASME codes for nuclear power plants and a standard 
for quality assurance. The NRC is also proposing to incorporate by 
reference four ASME code cases. As described in the ``Background'' and 
``Discussion'' sections of this notice, these materials provide rules 
for safety governing the design, fabrication, and inspection of nuclear 
power plant components.
    The NRC is required by law to obtain approval for incorporation by 
reference from the Office of the Federal Register (OFR). The OFR's 
requirements for incorporation by reference are set forth in 1 CFR part 
51. On November 7, 2014, the OFR adopted changes to its regulations 
governing incorporation by reference (79 FR 66267). The OFR regulations 
require an agency to include in a proposed rule a discussion of the 
ways that the materials the agency proposes to incorporate by reference 
are reasonably available to interested parties or how it worked to make 
those materials reasonably available to interested parties. The 
discussion in this section complies with the requirement for proposed 
rules as set forth in 10 CFR 51.5(a)(1).
    The NRC considers ``interested parties'' to include all potential 
NRC stakeholders, not only the individuals and entities regulated or 
otherwise subject to the NRC's regulatory oversight. These NRC 
stakeholders are not a homogenous group but vary with respect to the 
considerations for determining reasonable availability. Therefore, the 
NRC distinguishes between different classes of interested parties for 
purposes of determining whether the material is ``reasonably 
available.'' The NRC considers the following to be classes of 
interested parties in NRC rulemakings with regard to the material to be 
incorporated by reference:
     Individuals and small entities regulated or otherwise 
subject to the NRC's regulatory oversight (this class also includes 
applicants and potential applicants for licenses and other NRC 
regulatory approvals) and who are subject to the material to be 
incorporated by reference by rulemaking. In this context, ``small 
entities'' has the same meaning as a ``small entity'' under 10 CFR 
2.810.
     Large entities otherwise subject to the NRC's regulatory 
oversight (this class also includes applicants and potential applicants 
for licenses and other NRC regulatory approvals) and who are subject to 
the material to be incorporated by reference by rulemaking. In this 
context, ``large entities'' are those which do not qualify as a ``small 
entity'' under 10 CFR 2.810.
     Non-governmental organizations with institutional 
interests in the matters regulated by the NRC.
     Other Federal agencies, states, local governmental bodies 
(within the meaning of 10 CFR 2.315(c)).
     Federally-recognized and State-recognized \3\ Indian 
tribes.
---------------------------------------------------------------------------

    \3\ State-recognized Indian tribes are not within the scope of 
10 CFR 2.315(c). However, for purposes of the NRC's compliance with 
1 CFR 51.5, ``interested parties'' includes a broad set of 
stakeholders, including State-recognized Indian tribes.
---------------------------------------------------------------------------

     Members of the general public (i.e., individual, 
unaffiliated members of the public who are not regulated or otherwise 
subject to the NRC's regulatory oversight) who may wish to gain access 
to the materials which the NRC proposes to incorporate by reference by 
rulemaking in order to participate in the rulemaking.
    The NRC makes the materials to be incorporated by reference 
available for inspection to all interested parties, by appointment, at 
the NRC Technical Library, which is located at Two White Flint North, 
11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-
7000; email: Library.Resource@nrc.gov.
    Interested parties may purchase a copy of the materials from ASME 
at Three Park Avenue, New York, NY 10016, or at the ASME Web site 
https://www.asme.org/shop/standards. The materials are also accessible 
through third-party subscription services such as IHS (15 Inverness Way 
East, Englewood, CO 80112; https://global.ihs.com) and Thomson Reuters 
Techstreet (3916 Ranchero Dr., Ann Arbor, MI 48108; https://www.techstreet.com). The purchase prices for individual documents range 
from $225 to $720 and the cost to purchase all documents is 
approximately $9,000.
    For the class of interested parties constituting members of the 
general public who wish to gain access to the materials to be 
incorporated by reference in order to participate in the rulemaking, 
the NRC recognizes that the $9,000 cost may be so high that the 
materials could be regarded as not reasonably available for purposes of 
commenting on this rulemaking, despite the NRC's actions to make the 
materials available at the NRC's PDR. Accordingly, the NRC sent a 
letter to the ASME requesting that they consider enhancing public 
access to these materials during the public comment period (ADAMS 
Accession No. ML15085A206). In an April 21, 2015, letter to the NRC, 
the ASME agreed to make the materials available online in a read-only 
electronic access format during the public comment period (ADAMS 
Accession No. ML15112A064). Therefore, the seven editions and addenda 
to the ASME codes for nuclear power plants, the ASME standard for 
quality assurance, and the four ASME code cases which the NRC proposes 
to incorporate by reference in this rulemaking are available in read-
only format at the ASME Web site https://go.asme.org/NRC.
    The NRC concludes that the materials the NRC proposes to 
incorporate by reference in this rulemaking are reasonably available to 
all interested parties because the materials are available to all 
interested parties in multiple ways and in a manner consistent with 
their interest in the materials.

X. Environmental Assessment and Final Finding of No Significant 
Environmental Impact

    This proposed rule action is in accordance with the NRC's policy to 
incorporate by reference in Sec.  50.55a new editions and addenda of 
the ASME BPV and OM Codes to provide updated rules for constructing and 
inspecting components and testing pumps, valves, and dynamic restraints 
(snubbers) in light-water nuclear power plants. The ASME Codes are 
national voluntary consensus standards and are required by the NTTAA to 
be used by government agencies unless the use of such a standard is 
inconsistent with applicable law or otherwise impractical. The National 
Environmental Policy Act (NEPA) requires Federal agencies to study the 
impacts of their ``major Federal actions significantly affecting the 
quality of the human environment,'' and prepare detailed statements on 
the environmental impacts of the proposed action and alternatives to 
the proposed action (42 U.S.C. Sec. 4332(C); NEPA Sec. 102(C)).
    The NRC has determined under NEPA, as amended, and the NRC's

[[Page 56849]]

regulations in subpart A of 10 CFR part 51, that this proposed rule is 
not a major Federal action significantly affecting the quality of the 
human environment and, therefore, an environmental impact statement is 
not required. The rulemaking does not significantly increase the 
probability or consequences of accidents, no changes are being made in 
the types of effluents that may be released off-site, and there is no 
significant increase in public radiation exposure. The NRC estimates 
the radiological dose to plant personnel performing the inspections 
required by ASME BPV Code Case N-770-2 would be about 3 rem per plant 
over a 10-year interval, and a one-time exposure for mitigating welds 
of about 30 rem per plant. The NRC estimates the radiological dose to 
plant personnel performing the inspections required by ASME BPV Code 
Case N-729-4 would be about 3 rem per plant over a 10-year interval and 
a one-time exposure for mitigating welds of about 30 rem per plant. As 
required by 10 CFR part 20, and in accordance with current plant 
procedures and radiation protection programs, plant radiation 
protection staff will continue monitoring dose rates and would make 
adjustments in shielding, access requirements, decontamination methods, 
and procedures as necessary to minimize the dose to workers. The 
increased occupational dose to individual workers stemming from the 
ASME BPV Code Case N-770-2 and N-729-4 inspections must be maintained 
within the limits of 10 CFR part 20 and as low as reasonably 
achievable. Therefore, the NRC concludes that the increase in 
occupational exposure would not be significant. The proposed rule does 
not involve non-radiological plant effluents and has no other 
environmental impact. Therefore, no significant non-radiological 
impacts are associated with this action. The determination of this 
environmental assessment is that there will be no significant off-site 
impact to the public from this action.

XI. Paperwork Reduction Act Statement

    This proposed rule contains new or amended collections of 
information subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 
3501 et seq.). This proposed rule has been submitted to the Office of 
Management and Budget for review and approval of the information 
collections.
    Type of submission, new or revision: Revision.
    The title of the information collection: Domestic Licensing of 
Production and Utilization Facilities: Incorporation by Reference of 
American Society of Mechanical Engineers Codes and Code Cases.
    The form number if applicable: Not applicable.
    How often the collection is required or requested: On occasion.
    Who will be required or asked to respond: Power reactor licensees 
and applicants for power reactors under construction.
    An estimate of the number of annual responses: 320.
    The estimated number of annual respondents: 104.
    An estimate of the total number of hours needed annually to comply 
with the information collection requirement or request: 121,600.
    Abstract: This proposed rule is the latest in a series of 
rulemakings to amend the NRC's regulations to incorporate by reference 
revised and updated ASME codes for nuclear power plants. The number of 
operating nuclear power plants has decreased and the NRC has increased 
its estimate of the burden associated with developing alternative 
requests. Overall, the reporting burden for 10 CFR 50.55a has 
increased.
    The U.S. Nuclear Regulatory Commission is seeking public comment on 
the potential impact of the information collections contained in this 
proposed rule and on the following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of the burden of the proposed information 
collection accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the proposed information collection on 
respondents be minimized, including the use of automated collection 
techniques or other forms of information technology?
    A copy of the OMB clearance package and proposed rule is available 
in ADAMS (Accession Nos. ML14141A281 and ML14258B191) or may be viewed 
free of charge at the NRC's PDR, One White Flint North, 11555 Rockville 
Pike, Room O-1 F21, Rockville, MD 20852. You may obtain information and 
comment submissions related to the OMB clearance package by searching 
on https://www.regulations.gov under Docket ID NRC-2011-0088.
    You may submit comments on any aspect of these proposed information 
collection(s), including suggestions for reducing the burden and on the 
previously stated issues, by the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2011-0088.
     Mail comments to: FOIA, Privacy, and Information 
Collections Branch, Office of Information Services, Mail Stop: T-5 F53, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or to 
Vlad Dorjets, Desk Officer, Office of Information and Regulatory 
Affairs (3150-0011), NEOB-10202, Office of Management and Budget, 
Washington, DC 20503; telephone 202-395-7315, email: 
oira_submission@omb.eop.gov.
    Submit comments by October 19, 2015. Comments received after this 
date will be considered if it is practical to do so, but the NRC staff 
is able to ensure consideration only for comments received on or before 
this date.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

XII. Regulatory Analysis: Availability

    The NRC has prepared a draft regulatory analysis on this proposed 
rule. The analysis examines the costs and benefits of the alternatives 
considered by the Commission. The NRC requests public comments on the 
draft regulatory analysis. Comments on the draft analysis may be 
submitted to the NRC by any method provided in the ADDRESSES section of 
this notice.

XIII. Backfitting and Issue Finality

Introduction

    The NRC's Backfit Rule in Sec.  50.109 states that the NRC shall 
require the backfitting of a facility only when it finds the action to 
be justified under specific standards stated in the rule. Section 
50.109(a)(1) defines backfitting as the modification of or addition to 
systems, structures, components, or design of a facility; the design 
approval or manufacturing license for a facility; or the procedures or 
organization required to design, construct, or operate a facility. Any 
of these modifications or additions may result from a new or amended 
provision in the NRC's rules or the imposition of a regulatory position 
interpreting the NRC's rules that is either new or different from a 
previously applicable NRC position after issuance of the construction 
permit

[[Page 56850]]

or the operating license or the design approval.
    Section 50.55a requires nuclear power plant licensees to:
     Construct ASME BPV Code Class 1, 2, and 3 components in 
accordance with the rules provided in Section III, Division 1, of the 
ASME BPV Code (``Section III'').
     Inspect Class 1, 2, 3, Class MC, and Class CC components 
in accordance with the rules provided in Section XI, Division 1, of the 
ASME BPV Code (``Section XI'').
     Test Class 1, 2, and 3 pumps, valves, and dynamic 
restraints (snubbers) in accordance with the rules provided in the ASME 
OM Code.
    This rulemaking proposes to incorporate by reference the 2009 
Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition to the ASME 
BPV Code, Section III, Division 1 and ASME BPV Code, Section XI, 
Division 1, including NQA-1 (with conditions on its use), as well as 
the 2009 Edition and 2011 Addenda and 2012 Edition to the ASME OM Code 
and Code Cases N-770-2 and N-729-4.
    The ASME BPV and OM codes are national consensus standards 
developed by participants with broad and varied interests, in which all 
interested parties (including the NRC and utilities) participate. A 
consensus process involving a wide range of stakeholders is consistent 
with the National Technology Transfer and Advancement Act, inasmuch as 
the NRC has determined that there are sound regulatory reasons for 
establishing regulatory requirements for design, maintenance, ISI, and 
IST by rulemaking. The process also facilitates early stakeholder 
consideration of backfitting issues. Thus, the NRC believes that the 
NRC need not address backfitting with respect to the NRC's general 
practice of incorporating by reference updated ASME Codes.

Overall Backfitting Considerations: Section III of the ASME BPV Code

    Incorporation by reference of more recent editions and addenda of 
Section III of the ASME BPV Code does not affect a plant that has 
received a construction permit or an operating license or a design that 
has been approved. This is because the edition and addenda to be used 
in constructing a plant are, under Sec.  50.55a, determined based on 
the date of the construction permit, and are not changed thereafter, 
except voluntarily by the licensee. The incorporation by reference of 
more recent editions and addenda of Section III ordinarily applies only 
to applicants after the effective date of the final rule incorporating 
these new editions and addenda. Thus, incorporation by reference of a 
more recent edition and addenda of Section III does not constitute 
``backfitting'' as defined in Sec.  50.109(a)(1).

Overall Backfitting Considerations: Section XI of the ASME BPV Code and 
the ASME OM Code

    Incorporation by reference of more recent editions and addenda of 
Section XI of the ASME BPV Code and the ASME OM Code affects the ISI 
and IST programs of operating reactors. However, the Backfit Rule 
generally does not apply to incorporation by reference of later 
editions and addenda of the ASME BPV Code (Section XI) and OM Code. As 
previously mentioned, the NRC's longstanding regulatory practice has 
been to incorporate later versions of the ASME Codes into Sec.  50.55a. 
Under Sec.  50.55a, licensees shall revise their ISI and IST programs 
every 120 months to the latest edition and addenda of Section XI of the 
ASME BPV Code and the ASME OM Code incorporated by reference into Sec.  
50.55a 12 months before the start of a new 120-month ISI and IST 
interval. Thus, when the NRC approves and requires the use of a later 
version of the Code for ISI and IST, it is implementing this 
longstanding regulatory practice and requirement.
    Other circumstances where the NRC does not apply the Backfit Rule 
to the approval and requirement to use later Code editions and addenda 
are as follows:
    1. When the NRC takes exception to a later ASME BPV Code or OM Code 
provision but merely retains the current existing requirement, 
prohibits the use of the later Code provision, limits the use of the 
later Code provision, or supplements the provisions in a later Code. 
The Backfit Rule does not apply because the NRC is not imposing new 
requirements. However, the NRC explains any such exceptions to the Code 
in the Statement of Considerations and regulatory analysis for the 
rule.
    2. When an NRC exception relaxes an existing ASME BPV Code or OM 
Code provision but does not prohibit a licensee from using the existing 
Code provision. The Backfit Rule does not apply because the NRC is not 
imposing new requirements.
    3. Modifications and limitations imposed during previous routine 
updates of Sec.  50.55a have established a precedent for determining 
which modifications or limitations are backfits, or require a backfit 
analysis (e.g., final rule dated September 10, 2008 [73 FR 52731], and 
a correction dated October 2, 2008 [73 FR 57235]). The application of 
the backfit requirements to modifications and limitations in the 
current rule are consistent with the application of backfit 
requirements to modifications and limitations in previous rules.
    The incorporation by reference and adoption of a requirement 
mandating the use of a later ASME BPV Code or OM Code may constitute 
backfitting in some circumstances. In these cases, the NRC would 
perform a backfit analysis or documented evaluation in accordance with 
Sec.  50.109. These include the following:
    1. When the NRC endorses a later provision of the ASME BPV Code or 
OM Code that takes a substantially different direction from the 
existing requirements, the action is treated as a backfit (e.g., 61 FR 
41303 [August 8, 1996]).
    2. When the NRC requires implementation of a later ASME BPV Code or 
OM Code provision on an expedited basis, the action is treated as a 
backfit. This applies when implementation is required sooner than it 
would be required if the NRC simply endorsed the Code without any 
expedited language (e.g., 64 FR 51370 [September 22, 1999]).
    3. When the NRC takes an exception to an ASME BPV Code or OM Code 
provision and imposes a requirement that is substantially different 
from the existing requirement as well as substantially different from 
the later Code (e.g., 67 FR 60529 [September 26, 2002]).

Detailed Backfitting Discussion: Proposed Changes Beyond Those 
Necessary To Incorporate by Reference the New ASME BPV and OM Code 
Provisions

    This section discusses the backfitting considerations for all the 
proposed changes to Sec.  50.55a that go beyond the minimum changes 
necessary and required to adopt the new ASME Code Addenda into Sec.  
50.55a.

ASME BPV Code, Section III

    1. Revise Sec.  50.55a(b)(1)(ii), ``Weld leg dimensions,'' to 
clarify rule language and add Table 1, which clarifies prohibited 
Section III provisions in tabular form for welds with leg size less 
than 1.09 tn. This proposed change would not alter the 
original intent of this requirement and, therefore, would not impose a 
new requirement. Therefore, this proposed change is not a backfit.
    2. Revise Sec.  50.55a(b)(1)(iv), ``Section III condition: Quality 
assurance,'' to require that when applying editions and addenda later 
than the 1989 Edition of

[[Page 56851]]

Section III, the requirements of NQA-1, 1983 Edition through the 1994 
Edition, 2008 Edition, and the 2009-1a Addenda are acceptable for use, 
provided that the edition and addenda of NQA-1 specified in either NCA-
4000 or NCA-7000 is used in conjunction with the administrative, 
quality and technical provisions contained in the edition and addenda 
of Section III being used. This proposed revision clarifies the current 
requirements, and is considered to be consistent with the meaning and 
intent of the current requirements, and therefore is not considered to 
result in a change in requirements. Therefore, this proposed change is 
not a backfit.
    3. Add a new proposed condition as Sec.  50.55a(b)(1)(viii), ``Use 
of ASME Certification Marks,'' to allow licensees to use either the 
ASME BPV Code Symbol Stamp or ASME Certification Mark with the 
appropriate certification designator and class designator as specified 
in the 2013 Edition through the latest edition and addenda incorporated 
by reference in 10 CFR 50.55a. This proposed condition would not result 
in a change in requirements previously approved in the Code and, 
therefore, is not a backfit.

ASME BPV Code, Section XI

    1. Revise Sec.  50.55a(b)(2)(vi), ``Effective Edition and Addenda 
of Subsection IWE and Subsection IWL, Section XI;'' to clarify that the 
provision applies only to the class of licensees of operating reactors 
that were required by previous versions of Sec.  50.55a to develop, 
implement a containment inservice inspection program in accordance with 
Subsection IWE and Subsection IWL, and complete an expedited 
examination of containment during the 5-year period from September 9, 
1996, to September 9, 2001. This proposed revision clarifies the 
current requirements, is considered to be consistent with the meaning 
and intent of the current requirements, and is not considered to result 
in a change in requirements. Therefore, this proposed change is not a 
backfit.
    2. Revise Sec.  50.55a(b)(2)(viii), ``Examination of concrete 
containments,'' so that when using the 2007 Edition with 2009 Addenda 
through the 2013 Edition of Subsection IWL, the conditions in 10 CFR 
50.55a(b)(2)(viii)(E) do not apply, but the proposed conditions in new 
10 CFR 50.55a(b)(2)(viii)(H) and 10 CFR 50.55a(b)(2)(viii)(I) do apply. 
This proposed revision would not require 10 CFR 50.55a(b)(2)(viii)(E) 
to be used when following the 2007 Edition with 2009 Addenda through 
the 2013 Edition of Subsection IWL because most of its requirements 
have been included in IWL-2512, ``Inaccessible Areas.'' Therefore, this 
proposed change is not a backfit because the requirements have not 
changed. The revision to add the condition in 10 CFR 
50.55a(b)(2)(viii)(H) captures the reporting requirements of the 
current 10 CFR 50.55a(b)(2)(viii)(E) which were not included in IWL-
2512. Therefore, this proposed change is not a backfit because the 
requirements have not changed. The revision to add the condition in 10 
CFR 50.55a(b)(2)(viii)(I) addresses a new code provision in IWL-2512(b) 
for evaluation of below-grade concrete surfaces during the period of 
extended operation of a renewed license. The condition assures 
consistency with the GALL Report and applies to plants going forward 
using the 2007 Edition with 2009 Addenda through the 2013 Edition of 
Subsection IWL. The requirements would remain unchanged from those of 
the GALL Report and, therefore, this change is not a backfit.
    3. Revise Sec.  50.55a(b)(2)(ix), ``Examination of metal 
containments,'' to extend the applicability of the existing conditions 
in Sec.  50.55a(b)(2)(ix)(A)(2), Sec.  50.55a(b)(2)(ix)(B), and Sec.  
50.55a(b)(2)(ix)(J) to the 2007 Edition with 2009 Addenda through the 
2013 Edition of Subsection IWE. This proposed condition would not 
result in a change to current requirements, and is therefore not a 
backfit.
    4. Revise Sec.  50.55a(b)(2)(x), ``Section XI condition: Quality 
assurance,'' to require that when applying the editions and addenda 
later than the 1989 Edition of ASME BPV Code, Section XI, the 
requirements of NQA-1, 1983 Edition through the 1994 Edition, the 2008 
Edition, and the 2009-1a Addenda specified in either IWA-1400 or Table 
IWA 1600-1, ``Referenced Standards and Specifications,'' of that 
edition and addenda of Section XI are acceptable for use, provided the 
licensee uses its appendix B to 10 CFR part 50 quality assurance 
program in conjunction with Section XI requirements. This proposed 
revision clarifies the current requirements, which the NRC considers to 
be consistent with the meaning and intent of the current requirements. 
Therefore, the NRC does not consider the clarification to be a change 
in requirements. Therefore, this proposed change is not a backfit.
    5. Add a new proposed condition as Sec.  50.55a(b)(2)(xviii)(D), 
``NDE personnel certification: Fourth provision;'' to prohibit the use 
of Appendix VII and subarticle VIII-2200 of the 2011 Addenda and 2013 
Edition of Section XI of the ASME BPV Code. Licensees would be required 
to implement Appendix VII and subarticle VIII-2200 of the 2010 Edition 
of Section XI. This condition does not constitute a change in NRC 
position because the use of the subject provisions is not currently 
allowed by Sec.  50.55a. Therefore, the addition of this new proposed 
condition is not a backfit.
    6. Revise Sec.  50.55a(b)(2)(xxi)(A), ``Table IWB-2500-1 
examination requirements; First provision,'' to modify the standard for 
visual magnification resolution sensitivity and contrast for visual 
examinations of Examination Category B-D components, making the rule 
conform with ASME BPV Code, Section XI requirements for VT-1 
examinations. This proposed revision removes a condition that was in 
addition to the ASME Code requirements and does not impose a new 
requirement. Therefore, this change is not a backfit.
    7. Add a new proposed condition as Sec.  50.55a(b)(2)(xxx), ``Steam 
Generator Preservice Examinations;'' to require that instead of the 
preservice inspection requirements of Section XI, IWB-2200(c), a full 
length examination of 100 percent of the tubing in each newly installed 
steam generator shall be performed prior to plant startup. This 
proposed condition provides a clarification consistent with industry 
guidelines and the NRC staff position in SRP Section 5.4.2.2. 
Therefore, the addition of this new proposed condition is not a 
backfit.
    8. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxi), 
``Mechanical clamping devices;'' to prohibit the use of mechanical 
clamping devices in accordance with IWA-4131.1(c) in the 2010 Edition 
and IWA-4131.1(d) in the 2011 Addenda through 2013 Edition on small 
item Class 1 piping and portions of a piping system that forms the 
containment boundary. This condition does not constitute a change in 
NRC position and would not affect licensees because the use of the 
subject provisions is not currently allowed by Sec.  50.55a. Therefore, 
the addition of this new proposed condition is not a backfit.
    9. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxii), 
``Summary Report submittal;'' to clarify that licensees using the 2010 
Edition or later editions and addenda of Section XI must continue to 
submit to the NRC the Preservice and Inservice Summary Reports required 
by IWA-6240 of the 2009 addenda of Section XI. This proposed condition 
would not result in a change in NRC's requirements insomuch as these 
reports have been required in the 2009 Addenda of Section XI and all 
previous editions and

[[Page 56852]]

addenda. Therefore, the addition of this new proposed condition is not 
a backfit.
    10. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxiii), 
``Risk-Informed allowable pressure;'' to prohibit the use of ASME BPV 
Code, Section XI, Appendix G, Paragraph G-2216. The use of Paragraph G-
2216 is not currently allowed by Sec.  50.55a. Therefore, the proposed 
condition does not constitute a new or changed NRC position on the lack 
of acceptability of Paragraph G-2216. Therefore, the addition of this 
new proposed condition is not a backfit.
    11. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxiv), 
``Disposition of flaws in Class 3 components;'' to require that when 
using the 2013 Edition of the ASME BPV Code, Section XI, the licensee 
shall use the acceptance standards of IWD-3510 for the disposition of 
flaws in Category D-A components. The condition is imposed to provide 
clarification and consistency in requirements between IWD-3410 and IWD-
3510. This proposed change would not alter the original intent of this 
requirement and, therefore, would not impose a new requirement. This 
proposed change is not a backfit.
    12. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxv), ``Use 
of RTT0 in the KIa and KIc 
equations;'' to specify that when licensees use ASME BPV Code, Section 
XI 2013 Edition Nonmandatory Appendix A paragraph A-4200, if 
T0 is available, then RTT0 may be used in place 
of RTNDT for applications using the KIc equation 
and the associated KIc curve, but not for applications using 
the KIa equation and the associated KIa curve. 
Conditions on the use of ASME BPV Code, Section XI, Nonmandatory 
Appendices do not constitute backfitting inasmuch as those provisions 
apply to voluntary actions initiated by the licensee to use the 
``nonmandatory compliance'' provisions in these Appendices of the 
proposed rule.
    13. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxvi), 
``Fracture toughness of irradiated materials;'' to require licensees 
using ASME BPV Code, Section XI 2013 Edition Nonmandatory Appendix A 
paragraph A-4400, to obtain NRC approval before using irradiated 
T0 and the associated RTT0 in establishing 
fracture toughness of irradiated materials. Conditions on the use of 
ASME BPV Code, Section XI, Nonmandatory Appendices do not constitute 
backfitting inasmuch as those provisions apply to voluntary actions 
initiated by the licensee to use the ``nonmandatory compliance'' 
provisions in these Appendices of the proposed rule.
    14. Add a new proposed condition as Sec.  50.55a(b)(2)(xxxvii), 
ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast Austenitic 
Piping Welds From the Outside Surface Section XI, Division 1,'' to 
allow the use of the code case as conditioned. Conditions on the use of 
ASME BPV Code Case N-824 do not constitute backfitting, inasmuch as the 
use of this code case is not required by the NRC but instead is an 
alternative which may be voluntarily used by the licensee (i.e., a 
``voluntary alternative'').

ASME OM Code

    1. Add a new proposed condition as Sec.  50.55a(b)(3)(ii)(A) to 
require that licensees evaluate the adequacy of the diagnostic test 
interval for each MOV and adjust the interval as necessary, but not 
later than 5 years or three refueling outages (whichever is longer) 
from initial implementation of Appendix III of the ASME OM Code. This 
proposed condition represents an exception to a later OM Code provision 
but merely retains the current NRC requirement in RG 1.192, and is 
therefore not a backfit because the NRC is not imposing a new 
requirement.
    2. Add a new proposed condition as Sec.  50.55a(b)(3)(ii)(B) to 
require that licensees ensure that the potential increase in core 
damage frequency and large early release frequency associated with the 
extension is acceptably small when extending exercise test intervals 
for high risk MOVs beyond a quarterly frequency. This proposed 
condition represents an exception to a later OM Code provision but 
merely retains the current NRC requirement in RG 1.192, and is 
therefore not a backfit because the NRC is not imposing a new 
requirement.
    3. Add a new proposed condition as Sec.  50.55a(b)(3)(ii)(C) to 
require, when applying Appendix III to the ASME OM Code, that licensees 
categorize MOVs according to their safety significance using the 
methodology described in ASME OM Code Case OMN-3 subject to the 
conditions discussed in RG 1.192, or using an MOV risk ranking 
methodology accepted by the NRC on a plant-specific or industry-wide 
basis in accordance with the conditions in the applicable safety 
evaluation. This proposed condition represents an exception to a later 
OM Code provision but merely retains the current NRC requirement in RG 
1.192, and is therefore not a backfit because the NRC is not imposing a 
new requirement.
    4. Add a new proposed condition as Sec.  50.55a(b)(3)(ii)(D) to 
require that, when applying Paragraph III-3600, ``MOV Exercising 
Requirements,'' of Appendix III to the OM Code, licensees shall verify 
that the stroke time of the MOV satisfies the assumptions in the plant 
safety analyses. This proposed condition retains the MOV stroke time 
requirement that was specified in previous editions and addenda of the 
ASME OM Code. The retention of this requirement is not a backfit.
    5. Add new proposed conditions as Sec.  50.55a(b)(3)(iii)(A) 
through Sec.  50.55a(b)(3)(iii)(D), ``OM condition: New Reactors;'' to 
apply specific conditions for IST programs applicable to licensees of 
new nuclear power plants in addition to the provisions of the ASME OM 
Code as incorporated by reference with conditions in Sec.  50.55a. 
Licensees of ``new reactors'' are, as identified in the proposed 
paragraph: (i) Holders of operating licenses for nuclear power reactors 
that received construction permits under this part on or after the date 
12 months after the effective date of this rulemaking and (ii) holders 
of COLs issued under 10 CFR part 52, whose initial fuel loading occurs 
on or after the date 12 months after the effective date of this 
rulemaking. This implementation schedule for new reactors is consistent 
with the NRC regulations in Sec.  50.55a(f)(4)(i). These proposed 
conditions represent an exception to a later OM Code provision but 
merely retain the current NRC requirement, and are therefore not a 
backfit because the NRC is not imposing a new requirement.
    6. Revise Sec.  50.55a(b)(3)(iv), ``OM condition: Check valves 
(Appendix II),'' to specify that Appendix II, ``Check Valve Condition 
Monitoring Program,'' of the OM Code, 2003 Addenda through the 2012 
Edition, is acceptable for use without conditions with the 
clarifications that (1) the maximum test interval allowed by Appendix 
II for individual check valves in a group of two valves or more must be 
supported by periodic testing of a sample of check valves in the group 
during the allowed interval and (2) the periodic testing plan must be 
designed to test each valve of a group at approximate equal intervals 
not to exceed the maximum requirement interval. The regulation is being 
revised to extend the applicability of this existing NRC condition on 
the OM Code to the 2012 Edition of the OM Code. This does not represent 
a change in the NRC's position that the condition is needed with 
respect to the OM Code. Therefore, this proposed condition is not a 
backfit.
    7. Add a new proposed condition as Sec.  50.55a(b)(3)(vii), ``OM 
condition: Subsection ISTB;'' to prohibit the use of Subsection ISTB in 
the 2011 Addenda to the ASME OM Code because the complete set of 
planned Code

[[Page 56853]]

modifications to support the changes to the comprehensive pump test 
acceptance criteria was not made in that addenda. This proposed 
condition represents an exception to a later OM Code provision but 
merely limits the use of the later Code provision, and is therefore not 
a backfit because the NRC is not imposing a new requirement.
    8. Add a new proposed condition as Sec.  50.55a(b)(3)(viii), ``OM 
condition: Subsection ISTE;'' to allow licensees to implement 
Subsection ISTE, ``Risk-Informed Inservice Testing of Components in 
Light-Water Reactor Nuclear Power Plants,'' in the ASME OM Code, 2009 
Edition, 2011 Addenda and 2012 Edition, where the licensee has obtained 
authorization to implement Subsection ISTE as an alternative to the 
applicable IST requirements in the ASME OM Code on a case-by-case basis 
in accordance with Sec.  50.55a(z). This proposed condition represents 
an exception to a later OM Code provision but merely limits the use of 
the later Code provision, and is therefore not a backfit because the 
NRC is not imposing a new requirement.
    9. Add a new proposed condition as Sec.  50.55a(b)(3)(ix), ``OM 
Condition: Subsection ISTF;'' to specify that licensees applying 
Subsection ISTF, 2012 Edition, shall satisfy the requirements of 
Mandatory Appendix V, ``Pump Periodic Verification Test Program,'' of 
the ASME OM Code, 2012 Edition. The proposed condition also specifies 
that Subsection ISTF, 2011 Addenda, is not acceptable for use. This 
proposed condition represents an exception to a later OM Code provision 
but merely limits the use of the later Code provision, and is therefore 
not a backfit because the NRC is not imposing a new requirement.
    10. Add a new proposed condition as Sec.  50.55a(b)(3)(x), ``OM 
condition: ASME OM Code Case OMN-20,'' to allow licensees to implement 
ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' in the ASME OM 
Code, 2012 Edition. This proposed condition allows voluntary action 
initiated by the licensee to use the code case and is, therefore, not a 
backfit.
    11. Add a new proposed condition as Sec.  50.55a(b)(3)(xi), ``OM 
condition: Valve Position Indication,'' to specify that when 
implementing ASME OM Code, Subsection ISTC-3700, ``Position 
Verification Testing,'' licensees shall supplement the ASME OM Code 
provisions as necessary to verify that valve operation is accurately 
indicated. This proposed condition clarifies the current requirements, 
and is considered to be consistent with the meaning and intent of the 
current requirements, and therefore is not considered to result in a 
change in requirements. As such, this proposed condition is not a 
backfit.
    12. Revise Sec.  50.55a(f), ``Inservice testing requirements,'' to 
clarify that the ASME OM Code includes provisions for preservice 
testing of components as part of its overall provisions for IST 
programs. No expansion of IST program scope is intended by this 
clarification. This proposed condition would not result in a change in 
requirements previously approved in the Code and is, therefore, not a 
backfit.
    13. Revise Sec.  50.55a(f)(3)(iii)(A), ``Class 1 pumps and valves: 
First provision,'' to state that the paragraph is applicable to pumps 
and valves that are within the scope of the ASME OM Code. This will 
align the scope of pumps and valves for inservice testing with the 
scope defined in the ASME OM Code and in SRP Section 3.9.6. This 
proposed condition would not result in a change in requirements 
previously approved in the Code and is, therefore, not a backfit.
    14. Revise Sec.  50.55a(f)(3)(iii)(B), ``Class 1 pumps and valves: 
Second provision,'' to state that the paragraph is applicable to pumps 
and valves that are within the scope of the ASME OM Code. This will 
align the scope of pumps and valves for inservice testing with the 
scope defined in the ASME OM Code and in SRP Section 3.9.6. This 
proposed condition would not result in a change in requirements 
previously approved in the Code and is, therefore, not a backfit.
    15. Revise Sec.  50.55a(f)(3)(iv)(A), ``Class 2 and 3 pumps and 
valves: First provision;'' to state that the paragraph is applicable to 
pumps and valves that are within the scope of the ASME OM Code and not 
covered by paragraph (f)(3)(iii)(A) for Class 1 pumps and valves. This 
will align the scope of pumps and valves for inservice testing with the 
scope defined in the ASME OM Code and in SRP Section 3.9.6. This 
proposed condition would not result in a change in requirements 
previously approved in the Code and is, therefore, not a backfit.
    16. Revise Sec.  50.55a(f)(3)(iv)(B), ``Class 2 and 3 pumps and 
valves: Second provision,'' to state that the paragraph is applicable 
to pumps and valves that are within the scope of the ASME OM Code and 
not covered by paragraph (f)(3)(iii)(B) for Class 1 pumps and valves. 
This will align the scope of pumps and valves for inservice testing 
with the scope defined in the ASME OM Code and in SRP Section 3.9.6. 
This proposed condition would not result in a change in requirements 
previously approved in the Code, and is therefore not a backfit.
    17. Revise Sec.  50.55a(f)(4), ``Inservice testing standards for 
operating plants;'' to state that the paragraph is applicable to pumps 
and valves that are within the scope of the ASME OM Code. This will 
align the scope of pumps and valves for inservice testing with the 
scope defined in the ASME OM Code and in SRP Section 3.9.6. This 
proposed condition would not result in a change in requirements 
previously approved in the Code, and is therefore not a backfit.

ASME BPV Code Case N-729-4

    Revise Sec.  50.55a(g)(6)(ii)(D), ``Reactor vessel head 
inspections'':
    On June 22, 2012, the ASME approved the fourth revision of ASME BPV 
Code Case N-729, (N-729-4). The NRC proposes to update the requirements 
of Sec.  50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV 
Code Case N-729-4, with conditions. The ASME BPV Code Case N-729-4 
contains similar requirements as N-729-1; however, N-729-4 also 
contains new requirements to address previous NRC conditions, including 
changes to inspection frequency and qualifications. The new NRC 
conditions on the use of ASME BPV Code Case N-729-4 address operational 
experience, clarification of implementation, and the use of 
alternatives to the code case.
    The current regulatory requirements for the examination of 
pressurized water reactor upper RPV heads that use nickel-alloy 
materials are provided in Sec.  50.55a(g)(6)(ii)(D). This section was 
first created by rulemaking, dated September 10, 2008, (73 FR 52730) to 
require licensees to implement ASME BPV Code Case N-729-1, with 
conditions, instead of the inspections previously required by the ASME 
BPV Code, Section XI. The action did constitute a backfit; however, NRC 
concluded that imposition of ASME BPV Code Case N-729-1, as 
conditioned, constituted an adequate protection backfit.
    The GDC for nuclear power plants (appendix A to 10 CFR part 50) or, 
as appropriate, similar requirements in the licensing basis for a 
reactor facility, provide bases and requirements for NRC assessment of 
the potential for, and consequences of, degradation of the reactor 
coolant pressure boundary (RCPB). The applicable GDC include GDC 14 
(Reactor Coolant Pressure Boundary), GDC 31 (Fracture Prevention of 
Reactor Coolant Pressure Boundary), and GDC 32 (Inspection of Reactor 
Coolant Pressure Boundary). General Design Criterion 14 specifies that 
the RCPB be designed, fabricated, erected, and tested so as to have an 
extremely low probability of abnormal leakage, of

[[Page 56854]]

rapidly propagating failure, and of gross rupture. General Design 
Criterion 31 specifies that the probability of rapidly propagating 
fracture of the RCPB be minimized. General Design Criterion 32 
specifies that components that are part of the RCPB have the capability 
of being periodically inspected to assess their structural and leak 
tight integrity.
    The NRC concludes that ASME BPV Code Case N-729-4, as conditioned, 
shall be mandatory in order to ensure that the requirements of the GDC 
are satisfied. Imposition of ASME BPV Code Case N-729-4, with 
conditions, ensures that the ASME Code-allowable limits will not be 
exceeded, leakage will likely not occur and potential flaws will be 
detected before they challenge the structural or leak tight integrity 
of the reactor pressure vessel upper head within current nondestructive 
examination limitations. The NRC concludes that the regulatory 
framework for providing adequate protection of public health and safety 
is accomplished by the incorporation of ASME BPV Code Case N-729-4 into 
Sec.  50.55a, as conditioned. All current licensees of U.S. pressurized 
water reactors will be required to implement ASME BPV Code Case N-729-
4, as conditioned. The Code Case provisions on examination requirements 
for reactor pressure vessel upper heads are essentially the same as 
those established under ASME BPV Code Case N-729-1, as conditioned. One 
exception is the condition in Sec.  50.55a(g)(6)(ii)(D)(3), which will 
require, for upper heads with Alloy 600 penetration nozzles, that bare 
metal visual examinations be performed each outage in accordance with 
Table 1 of ASME BPV Code Case N-729-4. Accordingly, the NRC imposition 
of the ASME BPV Code Case N-729-4, as conditioned, may be deemed to be 
a modification of the procedures to operate a facility resulting from 
the imposition of the new regulation, and as such, this rulemaking 
provision may be considered backfitting under Sec.  50.109(a)(1).
    The NRC continues to find that inspections of reactor pressure 
vessel upper heads, their penetration nozzles, and associated partial 
penetration welds are necessary for adequate protection of public 
health and safety and that the requirements of ASME BPV Code Case N-
729-4, as conditioned, represent an acceptable approach, developed, in 
part, by a voluntary consensus standards organization for performing 
future inspections. The NRC concludes that approval of ASME BPV Code 
Case N-729-4, as conditioned, by incorporation by reference of the Code 
Case into Sec.  50.55a, is necessary to ensure that the facility 
provides adequate protection to the health and safety of the public and 
constitutes a redefinition of the requirements necessary to provide 
reasonable assurance of adequate protection of public health and 
safety. Therefore, a backfit analysis need not be prepared for this 
portion of the proposed rule in accordance with Sec.  50.109(a)(4)(ii) 
and Sec.  50.109(a)(4)(iii).

ASME BPV Code Case N-770-2

    Revise Sec.  50.55a(g)(6)(ii)(F), ``Examination requirements for 
Class 1 piping and nozzle dissimilar metal butt welds'':
    On June 9, 2011, the ASME approved the second revision of ASME BPV 
Code Case N-770, (N-770-2). The NRC proposes to update the requirements 
of Sec.  50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV 
Code Case N-770-2, with conditions. The ASME BPV Code Case N-770-2 
contains similar baseline and ISI requirements for unmitigated nickel-
alloy butt welds, and preservice and ISI requirements for mitigated 
butt welds as N-770-1. However, N-770-2 also contains new requirements 
for optimized weld overlays, a specific mitigation technique and 
volumetric inspection coverage. Further, the NRC conditions on the use 
of ASME BPV Code Case N-770-2 have been modified to address the changes 
in the code case, clarify inspection coverage requirements and require 
the development of inspection qualifications to allow complete weld 
inspection coverage in the future.
    The current regulatory requirements for the examination of ASME 
Class 1 piping and nozzle dissimilar metal butt welds that use nickel-
alloy materials is provided in Sec.  50.55a(g)(6)(ii)(F). This section 
was first created by rulemaking, dated June 21, 2011 (76 FR 36232), to 
require licensees to implement ASME BPV Code Case N-770-1, with 
conditions. The NRC added Sec.  50.55a(g)(6)(ii)(F) to require 
licensees to implement ASME BPV Code Case N-770-1, with conditions, 
instead of the inspections previously required by the ASME BPV Code, 
Section XI. The action did constitute a backfit; however, the NRC 
concluded that imposition of ASME BPV Code Case N-770-1, as 
conditioned, constituted an adequate protection backfit.
    The GDC for nuclear power plants (appendix A to 10 CFR part 50) or, 
as appropriate, similar requirements in the licensing basis for a 
reactor facility, provide bases and requirements for NRC assessment of 
the potential for, and consequences of, degradation of the RCPB. The 
applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC 
31 (Fracture Prevention of Reactor Coolant Pressure Boundary) and GDC 
32 (Inspection of Reactor Coolant Pressure Boundary). General Design 
Criterion 14 specifies that the RCPB be designed, fabricated, erected, 
and tested so as to have an extremely low probability of abnormal 
leakage, of rapidly propagating failure, and of gross rupture. General 
Design Criterion 31 specifies that the probability of rapidly 
propagating fracture of the RCPB be minimized. General Design Criterion 
32 specifies that components that are part of the RCPB have the 
capability of being periodically inspected to assess their structural 
and leak tight integrity.
    The NRC concludes that ASME BPV Code Case N-770-2, as conditioned, 
must be imposed in order to ensure that the requirements of the GDC are 
satisfied. Imposition of ASME BPV Code Case N-770-2, with conditions, 
ensures that the requirements of the GDC are met for all mitigation 
techniques currently in use for Alloy 82/182 butt welds because ASME 
Code-allowable limits will not be exceeded, leakage would likely not 
occur and potential flaws will be detected before they challenge the 
structural or leak tight integrity of piping welds. All current 
licensees of U.S. pressurized water reactors will be required to 
implement ASME BPV Code Case N-770-2, as conditioned. The Code Case 
provisions on examination requirements for ASME Class 1 piping and 
nozzle nickel-alloy dissimilar metal butt welds are somewhat different 
from those established under ASME BPV Code Case N-770-1, as 
conditioned, and will require a licensee to modify its procedures for 
inspection of ASME Class 1 nickel-alloy welds to meet these 
requirements. Accordingly, the NRC imposition of the ASME BPV Code Case 
N-770-2, as conditioned, may be deemed to be a modification of the 
procedures to operate a facility resulting from the imposition of the 
new regulation, and as such, this rulemaking provision may be 
considered backfitting under Sec.  50.109(a)(1).
    The NRC continues to find that ASME Class 1 nickel-alloy dissimilar 
metal weld inspections are necessary for adequate protection of public 
health and safety, and that the requirements of ASME BPV Code Case N-
770-2, as conditioned, represent an acceptable approach developed by a 
voluntary consensus standards organization for performing future ASME 
Class 1 nickel-alloy dissimilar metal weld inspections. The NRC 
concludes that approval of ASME BPV Code Case N-770-2, as conditioned, 
by incorporation by reference of the Code Case into Sec.  50.55a,

[[Page 56855]]

is necessary to ensure that the facility provides adequate protection 
to the health and safety of the public and constitutes a redefinition 
of the requirements necessary to provide reasonable assurance of 
adequate protection of public health and safety. Therefore, a backfit 
analysis need not be prepared for this portion of the proposed rule in 
accordance with Sec.  50.109(a)(4)(ii) and Sec.  50.109(a)(4)(iii).

Conclusion

    The NRC finds that incorporation by reference into Sec.  50.55a of 
the 2009 Addenda through 2013 Edition of Section III, Division 1, of 
the ASME BPV Code subject to the identified conditions; the 2009 
Addenda through 2013 Edition of Section XI, Division 1, of the ASME BPV 
Code, subject to the identified conditions; and the 2009 Edition 
through the 2012 Edition of the ASME OM Code subject to the identified 
conditions does not constitute backfitting or represent an 
inconsistency with any issue finality provisions in 10 CFR part 52.
    The NRC finds that the incorporation by reference of Code Cases N-
824 and OMN-20 does not constitute backfitting or represent an 
inconsistency with any issue finality provisions in 10 CFR part 52.
    The NRC finds that the inclusion of a new condition on Code Case N-
729-4 and a new condition on Code Case N-770-2 constitutes backfitting 
necessary for adequate protection.

XIV. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
NRC certifies that this proposed rule does not impose a significant 
economical impact on a substantial number of small entities. This 
proposed rule affects only the licensing and operation of commercial 
nuclear power plants. A licensee who is a subsidiary of a large entity 
does not qualify as a small entity. The companies that own these plants 
are not ``small entities'' as defined in the Regulatory Flexibility Act 
or the size standards established by the NRC (10 CFR 2.810), as the 
companies:
     Provide services that are not engaged in manufacturing, 
and have average gross receipts of more than $6.5 million over their 
last 3 completed fiscal years, and have more than 500 employees;
     Are not governments of a city, county, town, township or 
village;
     Are not school districts or special districts with 
populations of less than 50; and
     Are not small educational institutions.

XV. Availability of Documents

    The NRC is making the documents identified in Table 1 available to 
interested persons through one or more of the following methods, as 
indicated. To access documents related to this action, see the 
ADDRESSES section of this notice.

                   Table 1--Availability of Documents
------------------------------------------------------------------------
                  Document                       ADAMS Accession No.
------------------------------------------------------------------------
Proposed Rule Documents:
    Regulatory Analysis (includes            ML14170B104.
     backfitting discussion in Appendix A).
Related Documents:
    Fatigue and Fracture Mechanics: 33rd     ...........................
     Volume, ASTM STP 1417, W.G. Reuter and
     R.S. Piascik, Eds., ASTM
     International, West Conshohocken, PA,
     2002.
    ``Final Results from the CARINA Project  ...........................
     on Crack Initiation and Arrest of
     Irradiated German RPV Steels for
     Neutron Fluences in the Upper Bound,''
     by AREVA at the 26th Symposium on
     Effects of Radiation on Nuclear
     Materials (June 12-13, 2013,
     Indianapolis, IN, USA).
    Letter from Brian Thomas, NRC, to        ML15085A206.
     Michael Merker, ASME; ``Public Access
     to Material the NRC Seeks to
     Incorporate by Reference into its
     Regulations;'' April 9, 2015.
    Letter from Michael Merker, ASME, to     ML15112A064.
     Brian Thomas, NRC; April 21, 2015.
    Licensee Event Report 50-338/2012-001-   ML12151A441.
     00.
    NUREG-0800, ``Standard Review Plan for   ML070660036.
     the Review of Safety Analysis Reports
     for Nuclear Power Plants, LWR
     Edition''.
    NUREG-0800, Section 3.9.6, Revision 3,   ML070720041.
     ``Functional Design, Qualification,
     and Inservice Testing Programs for
     Pumps, Valves, and Dynamic
     Restraints,'' March 2007.
    NUREG-0800, Section 5.4.2.2, Revision    ML052340627.
     1, ``Steam Generator Tube Inservice
     Inspection,'' July 1981.
    NUREG-1482, Revision 2, ``Guidelines     ML13295A020.
     for Inservice Testing at Nuclear Power
     Plants: Inservice Testing of Pumps and
     Valves and Inservice Examination and
     Testing of Dynamic Restraints
     (Snubbers) at Nuclear Power Plants,''
     October 2013.
    NUREG-1801, Revision 2, ``Generic Aging  ML103490041.
     Lessons Learned (GALL) Report,''
     December 2010.
    NUREG-1950, ``Disposition of Public      ML11116A062.
     Comments and Technical Bases for
     Changes in the License Renewal
     Guidance Documents NUREG-1801 and
     NUREG-1800,'' April 2011.
    NUREG/CR-6860, ``An Assessment of        ML043630040.
     Visual Testing,'' November 2004.
    NUREG/CR-6933, ``Assessment of Crack     ML071020410 and
     Detection in Heavy-Walled Cast           ML071020414.
     Stainless Steel Piping Welds Using
     Advanced Low-Frequency Ultrasonic
     Methods,'' March 2007.
    NUREG/CR-7122, ``An Evaluation of        ML12087A004.
     Ultrasonic Phased Array Testing for
     Cast Austenitic Stainless Steel
     Pressurizer Surge Line Piping Welds,''
     March 2012.
    NRC Generic Letter 90-05, ``Guidance     ML031140590.
     for Performing Temporary Non-Code
     Repair of ASME Code Class 1, 2, and 3
     Piping (Generic Letter 90-05),'' June
     1990.
    NRC Meeting Summary of June 5-7, 2013,   ML14003A230.
     Annual Materials Programs Technical
     Information Exchange Public Meeting.
    NRC Memorandum, ``Consolidation of SECY- ML003708048.
     94-084 and SECY-95-132,'' July 24,
     1995.
    NRC Memorandum, ``Staff Requirements--   ML003755050.
     Affirmation Session, 11:30 a.m.,
     Friday, September 10, 1999,
     Commissioners' Conference Room, One
     White Flint North, Rockville, Maryland
     (Open to Public Attendance),''
     September 10, 1999.
    NRC Regulatory Guide 1.28, Revision 4,   ML100160003.
     ``Quality Assurance Program Criteria
     (Design and Construction),'' June 2010.
    NRC Regulatory Guide 1.83, Revision 1,   ML003740256.
     ``Inservice Inspection of Pressurized
     Water Reactor Steam Generator Tubes,''
     July 1975 (withdrawn in 2009).
    NRC Regulatory Guide 1.147, Revision     ML13339A689.
     17, ``Inservice Inspection Code Case
     Acceptability, ASME Section XI,
     Division 1,'' August 2014.

[[Page 56856]]

 
    NRC Regulatory Guide 1.174, Revision 2,  ML100910006.
     ``An Approach for Using Probabilistic
     Risk Assessment in Risk-Informed
     Decisions on Plant-Specific Changes to
     the Licensing Basis,'' May 2011.
    NRC Regulatory Guide 1.175, ``An         ML003740149.
     Approach for Plant-Specific, Risk-
     Informed Decisionmaking: Inservice
     Testing,'' August 1998.
    NRC Regulatory Guide 1.192, Revision 1,  ML13340A034.
     ``Operation and Maintenance Code Case
     Acceptability, ASME OM Code,'' August
     2014.
    NRC Regulatory Guide 1.200, Revision 2,  ML090410014.
     ``An Approach for Determining the
     Technical Adequacy of Probabilistic
     Risk Assessment Results for Risk-
     Informed Activities,'' March 2009.
    NRC Regulatory Guide 1.201, Revision 1,  ML061090627.
     ``Guidelines for Categorizing
     Structures, Systems, and Components in
     Nuclear Power Plants According to
     Their Safety Significance,'' May 2006.
    NRC Regulatory Information Conference,   https://www.nrc.gov/public-
     Recent Operating Reactors Materials      involve/conference-
     Issues, Presentation Materials, 2013.    symposia/ric/past/2013/
                                              docs/abstracts/
                                              sessionabstract-19.html.
    Relief Request REP-1 U2, Revision 2....  ML13232A308.
    SECY-90-016, ``Evolutionary Light Water  ML003707849.
     Reactor (LWR) Certification Issues and
     Their Relationship to Current
     Regulatory Requirements''.
    SECY-93-087, ``Policy, Technical, and    ML003708021.
     Licensing Issues Pertaining to
     Evolutionary and Advanced Light-Water
     Reactor (ALWR) Designs''.
    SECY-94-084, ``Policy and Technical      ML003708068.
     Issues Associated with the Regulatory
     Treatment of Non-Safety Systems in
     Passive Plant Designs''.
    SECY-95-132, ``Policy and Technical      ML003708005.
     Issues Associated with the Regulatory
     Treatment of Non-Safety Systems
     (RTNSS) in Passive Plant Designs (SECY-
     94-084)''.
ASME Codes, Standards, and Code Cases:
    ASME BPV Code, Section III, Division 1:  https://go.asme.org/NRC.
     2009 Addenda, 2010 Edition, 2011
     Addenda, and 2013 Edition.
    ASME BPV Code, Section XI, Division 1:   https://go.asme.org/NRC.
     2009 Addenda, 2010 Edition, 2011
     Addenda, and 2013 Edition.
    ASME OM Code, Division 1: 2009 Edition,  https://go.asme.org/NRC.
     2011 Addenda, and 2012 Edition.
    ASME Standard NQA-1: 1983 Edition        https://go.asme.org/NRC.
     through 1994 Edition, 2008 Edition,
     and 2009-1a Addenda.
    ASME BPV Code Case N-729-4.............  https://go.asme.org/NRC.
    ASME BPV Code Case N-770-2.............  https://go.asme.org/NRC.
    ASME BPV Code Case N-824...............  https://go.asme.org/NRC.
    ASME OM Code Case OMN-20...............  https://go.asme.org/NRC.
------------------------------------------------------------------------

    Throughout the development of this rulemaking, the NRC may post 
documents related to this rule, including public comments, on the 
Federal rulemaking Web site at https://www.regulations.gov under Docket 
ID NRC-2011-0088. The Federal rulemaking Web site allows you to receive 
alerts when changes or additions occur in a docket folder. To 
subscribe: (1) Navigate to the docket folder for NRC-2011-0088; (2) 
click the ``Sign up for Email Alerts'' link; and (3) enter your email 
address and select how frequently you would like to receive emails 
(daily, weekly, or monthly).

List of Subjects in 10 CFR Part 50

    Administrative practice and procedure, Antitrust, Classified 
information, Criminal penalties, Education, Fire prevention, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Penalties, Radiation protection, 
Reactor siting criteria, Reporting and recordkeeping requirements, 
Whistleblowing.

    For the reasons set forth in the preamble, and under the authority 
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553, the NRC proposes to adopt 
the following amendments to 10 CFR part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
1. The authority citation for part 50 continues to read as follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste 
Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National 
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 
note; Sec. 109, Public Law 96-295, 94 Stat. 783.

0
2. In Sec.  50.55a:
0
a. Revise paragraphs (a) introductory text, (a)(1)(i) introductory 
text, (a)(1)(i)(E)(12), (a)(1)(i)(E)(13) and add paragraphs 
(a)(1)(i)(E)(14) through (a)(1)(i)(E)(17);
0
b. Revise paragraph (a)(1)(ii) introductory text, (a)(1)(ii)(C)(48) and 
(a)(1)(ii)(C)(49) and add paragraphs (a)(1)(ii)(C)(50) through 
(a)(1)(ii)(C)(53);
0
c. Revise paragraphs (a)(1)(iii)(B) and (a)(1)(iii)(C) and add 
paragraphs (a)(1)(iii)(D), (a)(1)(iii)(E);
0
d. Revise paragraphs (a)(1)(iv) introductory text and add paragraphs 
(a)(1)(iv)(B) and (a)(1)(iv)(C);
0
e. Add paragraph (a)(1)(v);
0
f. Revise paragraphs (b) introductory text, (b)(1) introductory text, 
(b)(1)(ii), (b)(1)(iv), and (b)(1)(vii) and add paragraph (b)(1)(viii);
0
g. Revise paragraphs (b)(2) introductory text, (b)(2)(vi);
0
h. Revise paragraph (b)(2)(viii) introductory text and add paragraphs 
(b)(2)(viii)(H) and (b)(2)(viii)(I);
0
i. Revise paragraphs (b)(2)(ix) introductory text, (b)(2)(ix)(D), 
(b)(2)(x), add paragraph (b)(2)(xviii)(D), revise paragraph 
(b)(2)(xxi)(A), and add paragraphs (b)(2)(xxx) through (b)(2)(xxxvii);
0
j. Revise paragraphs (b)(3) introductory text, (b)(3)(i), and 
(b)(3)(ii), add paragraph (b)(3)(iii), revise paragraphs (b)(3)(iv) 
introductory text and (b)(3)(iv)(A) though (b)(3)(iv)(D), and add 
paragraphs (b)(3)(vii) through (b)(3)(xi);

[[Page 56857]]

0
k. Revise paragraphs (b)(4) introductory text, (b)(5), and (b)(6);
0
l. Revise paragraphs (f) introductory text, (f)(2), (f)(3)(iii)(A), 
(f)(3)(iii)(B), (f)(3)(iv)(A), (f)(3)(iv)(B), (f)(4) introductory text, 
(f)(4)(i), (f)(4)(ii);
0
m. Revise paragraphs (g) introductory text, (g)(2), (g)(3) introductory 
text, (g)(3)(i), (g)(3)(ii), (g)(3)(v), (g)(4)(i), (g)(4)(ii), and 
(g)(6)(ii)(D)(1) through (g)(6)(ii)(D)(4), remove paragraphs 
(g)(6)(ii)(D)(5) and (g)(6)(ii)(D)(6), revise paragraphs 
(g)(6)(ii)(F)(1) through (g)(6)(ii)(F)(10), and add paragraphs 
(g)(6)(ii)(F)(11) through (g)(6)(ii)(F)(13).
    The revisions and additions read as follows:


Sec.  50.55a  Codes and standards.

    (a) Documents approved for incorporation by reference. The 
standards listed in this paragraph have been approved for incorporation 
by reference by the Director of the Federal Register pursuant to 5 
U.S.C. 552(a) and 1 CFR part 51. The standards are available for 
inspection, by appointment, at the NRC Technical Library, which is 
located at Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland 20852; telephone: 301-415-7000; email: 
Library.Resource@nrc.gov; or at the National Archives and Records 
Administration (NARA). For information on the availability of this 
material at NARA, call 202-741-6030 or go to https://www.archives.gov/federal-register/cfr/ibr-locations.html.
    (1) * * *
    (i) ASME Boiler and Pressure Vessel Code, Section III. The editions 
and addenda for Section III of the ASME Boiler and Pressure Vessel Code 
(excluding Nonmandatory Appendices) are listed below, but limited by 
those provisions identified in paragraph (b)(1) of this section.
* * * * *
    (E) * * *
    (12) 2007 Edition,
    (13) 2008 Addenda,
    (14) 2009 Addenda,
    (15) 2010 Edition,
    (16) 2011 Addenda, and
    (17) 2013 Edition.
    (ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions 
and addenda for Section XI of the ASME Boiler and Pressure Vessel Code 
(excluding Nonmandatory Appendix U) are listed below, but limited by 
those provisions identified in paragraph (b)(2) of this section.
* * * * *
    (C) * * *
    (48) 2007 Edition,
    (49) 2008 Addenda,
    (50) 2009 Addenda,
    (51) 2010 Edition,
    (52) 2011 Addenda, and
    (53) 2013 Edition.
    (iii) * * *
    (B) ASME BPV Code Case N-729-4. ASME BPV Code Case N-729-4, 
``Alternative Examination Requirements for PWR Reactor Vessel Upper 
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds 
Section XI, Division 1'' (Approval Date: June 22, 2012), with the 
conditions in paragraph (g)(6)(ii)(D) of this section.
    (C) ASME BPV Code Case N-770-2. ASME BPV Code Case N-770-2, 
``Alternative Examination Requirements and Acceptance Standards for 
Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS 
N06082 or UNS W86182 Weld Filler Material With or Without Application 
of Listed Mitigation Activities Section XI, Division 1'' (Approval 
Date: June 9, 2011), with the conditions in paragraph (g)(6)(ii)(F) of 
this section.
    (D) ASME BPV Code Case N-824. ASME BPV Code Case N-824, 
``Ultrasonic Examination of Cast Austenitic Piping Welds From the 
Outside Surface Section XI, Division 1'' (Approval Date: October 16, 
2012), with the conditions in paragraphs (b)(2)(xxxvii)(A) through (E) 
of this section.
    (E) ASME OM Code Case OMN-20. ASME OM Code Case OMN-20, ``Inservice 
Test Frequency,'' in the 2012 Edition of the ASME OM Code. OMN-20 is 
referenced in paragraph (b)(3)(x).
    (iv) ASME Operation and Maintenance Code. The editions and addenda 
for the ASME Operation and Maintenance of Nuclear Power Plants are 
listed below, but limited by those provisions identified in paragraph 
(b)(3) of this section.
* * * * *
    (B) ``Operation and Maintenance of Nuclear Power Plants, Division 
1: Section IST Rules for Inservice Testing of Light-Water Reactor Power 
Plants''
    (1) 2009 Edition and
    (2) 2011 Addenda.
    (C) ``Operation and Maintenance of Nuclear Power Plants, Division 
1: OM Code: Section IST.''
    (1) 2012 Edition.
    (2) [Reserved]
    (v) ASME Quality Assurance Requirements.
    (A) ASME NQA-1, ``Quality Assurance Program Requirements for 
Nuclear Facilities.''
    (1) NQA-1-1983 Edition,
    (2) NQA-1a-1983 Addenda,
    (3) NQA-1b-1984 Addenda,
    (4) NQA-1c-1985 Addenda,
    (5) NQA-1-1986 Edition,
    (6) NQA-1a-1986 Addenda,
    (7) NQA-1b-1987 Addenda, and
    (8) NQA-1c-1988 Addenda.
    (9) NQA-1-1989 Edition,
    (10) NQA-1a-1989 Addenda,
    (11) NQA-1b-1991 Addenda, and
    (12) NQA-1c-1992 Addenda.
    (B) ASME NQA-1, ``Quality Assurance Requirements for Nuclear 
Facility Applications.''
    (1) NQA-1-1994 Edition,
    (2) NQA-1a-2008 Edition, and
    (3) NQA-1a-2009 Addenda.
* * * * *
    (b) Use and conditions on the use of standards. Systems and 
components of boiling and pressurized water-cooled nuclear power 
reactors must meet the requirements of the ASME Boiler and Pressure 
Vessel Code (BPV Code) and the ASME Operation and Maintenance of 
Nuclear Power Plants (OM Code) as specified in this paragraph. Each 
combined license for a utilization facility is subject to the following 
conditions.
    (1) Conditions on ASME BPV Code Section III. Each manufacturing 
license, standard design approval, and design certification under part 
52 of this chapter is subject to the following conditions. As used in 
this section, references to Section III refer to Section III of the 
ASME Boiler and Pressure Vessel Code and include the 1963 Edition 
through 1973 Winter Addenda and the 1974 Edition (Division 1) through 
the 2013 Edition (Division 1), subject to the following conditions:
* * * * *
    (ii) Section III condition: Weld leg dimensions. When applying the 
1989 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1) of this section, applicants and licensees 
may not apply the Section III provisions identified in Table 1 of this 
section for welds with leg size less than 1.09 tn

[[Page 56858]]



                              Table 1 of Sec.   50.55a--Prohibited Code Provisions
----------------------------------------------------------------------------------------------------------------
              Editions and Addenda                                        Code provision
----------------------------------------------------------------------------------------------------------------
1989 Addenda through 2013 Edition...............  Subparagraph NB-3683.4(c)(1).
                                                  Subparagraph NB-3683.4(c)(2).
1989 Addenda through 2003 Addenda...............  Note 11 to Figure NC-3673.2(b)-1.
                                                  Note 11 to Figure ND-3673.2(b)-1.
2004 Edition through 2010 Edition...............  Note 13 to Figure NC-3673.2(b)-1.
                                                  Note 13 to Figure ND-3673.2(b)-1.
2011 Addenda through 2013 Edition...............  Note 11 to Table NC-3673.2(b)-1.
                                                  Note 11 to Table ND-3673.2(b)-1.
----------------------------------------------------------------------------------------------------------------

* * * * *
    (iv) Section III condition: Quality assurance. When applying 
editions and addenda later than the 1989 Edition of Section III, the 
requirements of NQA-1, ``Quality Assurance Requirements for Nuclear 
Facility Applications,'' 1983 Edition through the 1994 Edition, 2008 
Edition, and the 2009-1a Addenda specified in either NCA-4000 or NCA-
7000 of that edition and addenda of Section III may be used by an 
applicant or licensee provided that the administrative, quality, and 
technical provisions contained in that edition and addenda of Section 
III are used in conjunction with the applicant's or licensee's appendix 
B to 10 CFR part 50 quality assurance program; and that commitments 
contained in the applicant's or licensee's quality assurance program 
description which are either more stringent than those contained in 
NQA-1 or have no comparable provision in NQA-1 or Section III, govern 
the applicant's or licensee's Section III activities.
* * * * *
    (vii) Section III condition: Capacity certification and 
demonstration of function of incompressible-fluid pressure-relief 
valves. When applying the 2006 Addenda through the 2013 Edition, 
applicants and licensees may use paragraph NB-7742, except that 
paragraph NB-7742(a)(2) may not be used. For a valve design of a single 
size to be certified over a range of set pressures, the demonstration 
of function tests under paragraph NB-7742 must be conducted as 
prescribed in NB-7732.2 on two valves covering the minimum set pressure 
for the design and the maximum set pressure that can be accommodated at 
the demonstration facility selected for the test.
    (viii) Section III condition: Use of ASME certification marks. When 
applying editions and addenda earlier than the 2011 Addenda to the 2010 
Edition, licensees may use either the ASME BPV Code Symbol Stamps or 
the ASME Certification Marks with the appropriate certification 
designators and class designators as specified in the 2013 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1) of this section.
    (2) Conditions on ASME BPV Code, Section XI. As used in this 
section, references to Section XI refer to Section XI, Division 1, of 
the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition 
through the 1976 Winter Addenda and the 1977 Edition through the 2013 
Edition (excluding Nonmandatory Appendix U), subject to the following 
conditions:
* * * * *
    (vi) Section XI condition: Effective edition and addenda of 
Subsection IWE and Subsection IWL. Licensees that implemented the 
expedited examination of containment, in accordance with Subsection IWE 
and Subsection IWL, during the period from September 9, 1996, to 
September 9, 2001, may use either the 1992 Edition with the 1992 
Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and 
Subsection IWL, as conditioned by the requirements in paragraphs 
(b)(2)(viii) and (ix) of this section, when implementing the initial 
120-month inspection interval for the containment inservice inspection 
requirements of this section. Successive 120-month interval updates 
must be implemented in accordance with paragraph (g)(4)(ii) of this 
section.
* * * * *
    (viii) Section XI condition: Concrete containment examinations. 
Applicants or licensees applying Subsection IWL, 1992 Edition with the 
1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through (E) of this 
section. Applicants or licensees applying Subsection IWL, 1995 Edition 
with the 1996 Addenda, must apply paragraphs (b)(2)(viii)(A), 
(b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of this section. Applicants or 
licensees applying Subsection IWL, 1998 Edition through the 2000 
Addenda, must apply paragraphs (b)(2)(viii)(E) and (F) of this section. 
Applicants or licensees applying Subsection IWL, 2001 Edition through 
the 2004 Edition, up to and including the 2006 Addenda, must apply 
paragraphs (b)(2)(viii)(E) through (G) of this section. Applicants or 
licensees applying Subsection IWL, 2007 Edition up to and including the 
2008 Addenda must apply paragraph (b)(2)(viii)(E) of this section. 
Applicants or licensees applying Subsection IWL, 2007 Edition with the 
2009 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section, must apply paragraph 
(b)(2)(viii)(H) and (b)(2)(viii)(I) of this section.
* * * * *
    (H) Concrete containment examinations: Eighth provision. For each 
inaccessible area of concrete identified for evaluation under IWL-2512, 
the licensee must provide the applicable information specified in 
paragraphs (b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and 
(b)(2)(viii)(E)(3) of this section in the ISI Summary Report required 
by IWA-6000.
    (I) Concrete containment examinations: Ninth provision. During the 
period of extended operation of a renewed license under part 54 of this 
chapter, the licensee must perform the technical evaluation under IWL-
2512(b) of inaccessible below-grade concrete surfaces exposed to 
foundation soil, backfill, or groundwater at periodic intervals not to 
exceed 5 years. In addition, the licensee must examine representative 
samples of the exposed portions of the below-grade concrete, when such 
below-grade concrete is excavated for any reason.
* * * * *
    (ix) Section XI condition: Metal containment examinations. 
Applicants or licensees applying Subsection IWE, 1992 Edition with the 
1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy 
the requirements of paragraphs (b)(2)(ix)(A) through (E) of this 
section. Applicants or licensees applying Subsection IWE, 1998 Edition 
through the 2001 Edition with the 2003 Addenda, must satisfy the 
requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) 
through (I) of this section. Applicants or licensees applying 
Subsection IWE, 2004 Edition, up to and including the 2005 Addenda, 
must

[[Page 56859]]

satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and 
(b)(2)(ix)(F) through (H) of this section. Applicants or licensees 
applying Subsection IWE, 2004 Edition with the 2006 Addenda, must 
satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and 
(b)(2)(ix)(B) of this section. Applicants or licensees applying 
Subsection IWE, 2007 Edition through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, must 
satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and 
(b)(2)(ix)(B) and (J) of this section.
* * * * *
    (D) Metal containment examinations: Fourth provision. This 
paragraph (b)(2)(ix)(D) may be used as an alternative to the 
requirements of IWE-2430. If the examinations reveal flaws or areas of 
degradation exceeding the acceptance standards of Table IWE-3410-1, an 
evaluation must be performed to determine whether additional component 
examinations are required. For each flaw or area of degradation 
identified that exceeds acceptance standards, the applicant or licensee 
must provide the following in the ISI Summary Report required by IWA-
6000:
    (1) A description of each flaw or area, including the extent of 
degradation, and the conditions that led to the degradation;
    (2) The acceptability of each flaw or area and the need for 
additional examinations to verify that similar degradation does not 
exist in similar components;
    (3) A description of necessary corrective actions; and
    (4) The number and type of additional examinations to ensure 
detection of similar degradation in similar components.
* * * * *
    (x) Section XI condition: Quality assurance. When applying the 
editions and addenda later than the 1989 Edition of ASME BPV Code, 
Section XI, the edition and addenda of NQA-1, ``Quality Assurance 
Requirements for Nuclear Facility Applications,'' 1983 Edition through 
the 1994 Edition, the 2008 Edition, and the 2009-1a Addenda specified 
in either IWA-1400 or Table IWA 1600-1 of that edition and addenda of 
Section XI, may be used by a licensee provided that the licensee uses 
its appendix B to 10 CFR part 50 quality assurance program in 
conjunction with Section XI requirements. Commitments contained in the 
licensee's quality assurance program description that are more 
stringent than those contained in NQA-1 must govern Section XI 
activities. Further, where NQA-1 and Section XI do not address the 
commitments contained in the licensee's appendix B quality assurance 
program description, the commitments must be applied to Section XI 
activities.
* * * * *
    (xviii) * * *
    (D) NDE personnel certification: Fourth provision. The use of 
Appendix VII and subarticle VIII-2200 of the 2011 Addenda and 2013 
Edition of Section XI of the ASME BPV Code is prohibited. When using 
ASME BPV Code, Section XI editions and addenda later than the 2010 
Edition, licensees and applicants must use the prerequisites for 
ultrasonic examination personnel certifications in Table VII-4110-1 and 
subarticle VIII-2200, Appendix VIII in the 2010 Edition.
* * * * *
    (xxi) * * *
    (A) Table IWB-2500-1 examination requirements: First provision. The 
provisions of Table IWB 2500-1, Examination Category B-D, Full 
Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 
(Inspection Program A) and Items B3.120 and B3.140 (Inspection Program 
B) of the 1998 Edition must be applied when using the 1999 Addenda 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1)(ii) of this section. A visual examination with 
magnification that has a resolution sensitivity to resolve 0.044 inch 
(1.1 mm) lower case characters without an ascender or descender (e.g., 
a, e, n, v), utilizing the allowable flaw length criteria in Table IWB-
3512-1, 1997 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, with 
a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may 
be performed instead of an ultrasonic examination.
* * * * *
    (xxx) Section XI condition: Steam generator preservice 
examinations. Prior to plant start up with a newly installed steam 
generator, a 100 percent full length examination will be conducted of 
the tubing in each new steam generator instead of the preservice 
inspection requirements of IWB-2200(c).
    (xxxi) Section XI condition: Mechanical clamping devices. The use 
of mechanical clamping devices on Class 1 piping and portions of piping 
systems that form the containment boundary is prohibited.
    (xxxii) Section XI condition: Summary report submittal. When using 
ASME BPV Code, Section XI, 2010 Edition through the latest edition and 
addenda incorporated by reference in paragraph (a)(1)(ii) of this 
section, Summary Reports described in IWA-6000 must be submitted to the 
NRC. Preservice inspection summary reports shall be submitted prior to 
the date of placement of the unit into commercial service and inservice 
inspection summary reports shall be submitted within 90 calendar days 
of the completion of each refueling outage.
    (xxxiii) Section XI condition: Risk-Informed allowable pressure. 
The use of Paragraph G-2216 in Appendix G in the 2011 Addenda and later 
editions and addenda of the ASME BPV Code, Section XI is prohibited.
    (xxxiv) Section XI condition: Disposition of flaws in Class 3 
components. When using the 2013 Edition of the ASME BPV Code, Section 
XI, to disposition flaws in Examination Category D-A components (i.e., 
welded attachments for vessels, piping, pumps, and valves), the 
acceptance standards of IWD-3510 must be used.
    (xxxv) Section XI condition: Use of RTT0 in the KIa and KIc 
equations. When using the 2013 Edition of the ASME BPV Code, Section 
XI, Appendix A, paragraph A-4200, if T0 is available, then 
RTT0 may be used in place of RTNDT for 
applications using the KIc equation and the associated 
KIc curve, but not for applications using the KIa 
equation and the associated KIa curve.
    (xxxvi) Section XI condition: Fracture toughness of irradiated 
materials. When using the 2013 Edition of the ASME BPV Code, Section 
XI, Appendix A paragraph A-4400, the licensee shall obtain NRC approval 
before using irradiated T0 and the associated 
RTT0 in establishing fracture toughness of irradiated 
materials.
    (xxxvii) Section XI condition: ASME BPV Code Case N-824. Licensees 
may use the provisions of ASME BPV Code Case N-824, ``Ultrasonic 
Examination of Cast Austenitic Piping Welds From the Outside Surface 
Section XI, Division 1,'' subject to the following conditions.
    (A) Ultrasonic examinations must be spatially encoded.
    (B) Instead of Paragraph 1(c)(1)(-a) licensees shall use dual, 
transmit-receive, refracted longitudinal wave, multi-element phased 
array search units.
    (C) Instead of Paragraph 1(c)(1)(-c) (-1), licensees shall use a 
phased array search unit with a center frequency between 500 kHz and 1 
MHz.
    (D) Instead of Paragraph 1(c)(1)(-c) (-2), licensees shall use a 
phased array search unit with a center frequency of 500 kHz.
    (E) Instead of Paragraph 1(c)(1)(-d), the phased array search unit 
must

[[Page 56860]]

produce angles from 30 to 70 degrees with a maximum increment of 5 
degrees.
    (3) Conditions on ASME OM Code. As used in this section, references 
to the OM Code are to the ASME OM Code, Subsections ISTA, ISTB, ISTC, 
ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and 
Nonmandatory Appendices A through H and J through M, in the 1995 
Edition through the 2012 Edition as specified in paragraph (a)(1)(iv). 
The following conditions are applicable when implementing the ASME OM 
Code:
    (i) OM condition: Quality assurance. When applying editions and 
addenda of the OM Code, the requirements of ASME Standard NQA-1, 
``Quality Assurance Requirements for Nuclear Facility Applications,'' 
1983 Edition through the 1994 Edition, 2008 Edition, and 2009-1a 
Addenda, are acceptable as permitted by either ISTA 1.4 of the 1995 
Edition through 1997 Addenda or ISTA-1500 of the 1998 Edition through 
the latest edition and addenda of the OM Code incorporated by reference 
in paragraph (a)(1)(iv) of this section, provided the licensee uses its 
appendix B to 10 CFR part 50 quality assurance program in conjunction 
with the OM Code requirements. Commitments contained in the licensee's 
quality assurance program description that are more stringent than 
those contained in NQA-1 govern OM Code activities. If NQA-1 and the OM 
Code do not address the commitments contained in the licensee's 
appendix B quality assurance program description, the commitments must 
be applied to OM Code activities.
    (ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees 
must comply with the provisions for testing MOVs in OM Code, ISTC 4.2, 
1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1)(iv) of this section, and must establish a program to 
ensure that MOVs continue to be capable of performing their design 
basis safety functions. Licensees implementing OM Code, Mandatory 
Appendix III, ``Preservice and Inservice Testing of Active Electric 
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,'' 
of the 2009 Edition, 2011 Addenda, and 2012 Edition shall comply with 
the following conditions:
    (A) MOV diagnostic test interval. Licensees shall evaluate the 
adequacy of the diagnostic test interval for each MOV and adjust the 
interval as necessary, but not later than 5 years or three refueling 
outages (whichever is longer) from initial implementation of OM Code, 
Appendix III.
    (B) MOV testing impact on risk. Licensees shall ensure that the 
potential increase in core damage frequency and large early release 
frequency associated with the extension is acceptably small when 
extending exercise test intervals for high risk MOVs beyond a quarterly 
frequency.
    (C) MOV risk categorization. When applying Appendix III to the OM 
Code, licensees shall categorize MOVs according to their safety 
significance using the methodology described in ASME OM Code Case OMN-
3, ``Requirements for Safety Significance Categorization of Components 
Using Risk Insights for Inservice Testing of LWR Power Plants,'' 
subject to the conditions applicable to OMN-3 which are set forth in 
Regulatory Guide 1.192, or using an MOV risk ranking methodology 
accepted by the NRC on a plant-specific or industry-wide basis in 
accordance with the conditions in the applicable safety evaluation.
    (D) MOV stroke time. When applying Paragraph III-3600, ``MOV 
Exercising Requirements,'' of Appendix III to the OM Code, licensees 
shall verify that the stroke time of the MOV satisfies the assumptions 
in the plant safety analyses.
    (iii) OM condition: New Reactors. In addition to complying with the 
provisions in the OM Code with the conditions specified in paragraph 
(b)(3) of this section, holders of operating licenses for nuclear power 
reactors that received construction permits under this part on or after 
the date 12 months after [the effective date of the final rule], and 
holders of combined licenses issued under 10 CFR part 52, whose initial 
fuel loading occurs on or after the date 12 months after [the effective 
date of the final rule] shall also comply with the following 
conditions, as applicable:
    (A) Power-operated valves. Licensees shall periodically verify the 
capability of power-operated valves to perform their design-basis 
safety functions.
    (B) Check valves. Licensees must perform bi-directional testing of 
check valves within the IST program where practicable.
    (C) Flow-induced vibration. Licensees shall monitor flow-induced 
vibration from hydrodynamic loads and acoustic resonance during 
preservice testing and inservice testing to identify potential adverse 
flow effects on components within the scope of the IST program.
    (D) High risk non-safety systems. Licensees shall assess the 
operational readiness of pumps, valves, and dynamic restraints within 
the scope of the Regulatory Treatment of Non-Safety Systems for 
applicable reactor designs.
    (iv) OM condition: Check valves (Appendix II). Appendix II, ``Check 
Valve Condition Monitoring Program,'' of the OM Code, 2003 Addenda 
through the 2012 Edition, is acceptable for use without conditions with 
the clarifications that (1) the maximum test interval allowed by 
Appendix II for individual check valves in a group of two valves or 
more must be supported by periodic testing of a sample of check valves 
in the group during the allowed interval and (2) the periodic testing 
plan must be designed to test each valve of a group at approximate 
equal intervals not to exceed the maximum requirement interval. 
Licensees applying Appendix II of the OM Code, 1995 Edition with the 
1996 and 1997 Addenda, shall satisfy the requirements of paragraphs 
(b)(3)(iv)(A) through (C) of this section. Licensees applying Appendix 
II, 1998 Edition through the 2012 Edition, shall satisfy the 
requirements of paragraphs (b)(3)(iv)(A), (B), and (D) of this section.
* * * * *
    (vii) OM condition: Subsection ISTB. Subsection ISTB, 2011 Addenda, 
is prohibited for use.
    (viii) OM condition: Subsection ISTE. Licensees may not implement 
the risk-informed approach for inservice testing (IST) of pumps and 
valves specified in Subsection ISTE, ``Risk-Informed Inservice Testing 
of Components in Light-Water Reactor Nuclear Power Plants,'' in the OM 
Code, 2009 Edition, 2011 Addenda, or 2012 Edition, without first 
obtaining NRC authorization to use Subsection ISTE as an alternative to 
the applicable IST requirements in the OM Code pursuant to Sec.  
50.55a(z).
    (ix) OM condition: Subsection ISTF. Licensees applying Subsection 
ISTF, 2012 Edition, shall satisfy the requirements of Mandatory 
Appendix V, ``Pump Periodic Verification Test Program,'' of the ASME OM 
Code, 2012 Edition. Subsection ISTF, 2011 Addenda, is not acceptable 
for use.
    (x) OM condition: ASME OM Code Case OMN-20. Licensees may implement 
ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' which is 
incorporated by reference in paragraph (a)(1)(iii)(E) of this section.
    (xi) OM condition: Valve Position Indication. When implementing 
ASME OM Code, Subsection ISTC-3700, ``Position Verification Testing,'' 
licensees shall develop and implement a method to verify that valve 
operation is accurately indicated by supplementing valve position 
indicating lights with other indications, such as flow meters or other 
suitable

[[Page 56861]]

instrumentation, to provide assurance of proper obturator position.
    (4) Conditions on Design, Fabrication, and Materials Code Cases. 
Each manufacturing license, standard design approval, and design 
certification application under part 52 of this chapter is subject to 
the following conditions. Licensees may apply the ASME BPV Code Cases 
listed in NRC Regulatory Guide 1.84, as incorporated by reference in 
paragraph (a)(3)(i) of this section, without prior NRC approval, 
subject to the following conditions:
* * * * *
    (5) Conditions on inservice inspection Code Cases. Licensees may 
apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporate by reference in paragraph (a)(3)(ii) of this section, 
without prior NRC approval, subject to the following:
    (i) ISI Code Case condition: Applying Code Cases. When a licensee 
initially applies a listed Code Case, the licensee must apply the most 
recent version of that Code Case incorporated by reference in paragraph 
(a) of this section.
    (ii) ISI Code Case condition: Applying different revisions of Code 
Cases. If a licensee has previously applied a Code Case and a later 
version of the Code Case is incorporated by reference in paragraph (a) 
of this section, the licensee may continue to apply, to the end of the 
current 120-month interval, the previous version of the Code Case, as 
authorized, or may apply the later version of the Code Case, including 
any NRC-specified conditions placed on its use. Licensees who choose to 
continue use of the Code Case during subsequent 120-month ISI program 
intervals will be required to implement the latest version incorporated 
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of NRC 
Regulatory Guide 1.147, as incorporated by reference in paragraph 
(a)(3)(ii) of this section.
    (iii) ISI Code Case condition: Applying annulled Code Cases. 
Application of an annulled Code Case is prohibited unless a licensee 
previously applied the listed Code Case prior to it being listed as 
annulled in NRC Regulatory Guide 1.147. If a licensee has applied a 
listed Code Case that is later listed as annulled in NRC Regulatory 
Guide 1.147, the licensee may continue to apply the Code Case to the 
end of the current 120-month interval.
    (6) Conditions on Operation and Maintenance of Nuclear Power Plants 
Code Cases. Licensees may apply the ASME Operation and Maintenance Code 
Cases listed in NRC Regulatory Guide 1.192, as incorporated by 
reference in paragraph (a)(3)(iii), without prior NRC approval, subject 
to the following:
    (i) OM Code Case condition: Applying Code Cases. When a licensee 
initially applies a listed Code Case, the licensee must apply the most 
recent version of that Code Case incorporated by reference in paragraph 
(a) of this section.
    (ii) OM Code Case condition: Applying different revisions of Code 
Cases. If a licensee has previously applied a Code Case and a later 
version of the Code Case is incorporated by reference in paragraph (a) 
of this section, the licensee may continue to apply, to the end of the 
current 120-month interval, the previous version of the Code Case, as 
authorized, or may apply the later version of the Code Case, including 
any NRC-specified conditions placed on its use. Licensees who choose to 
continue use of the Code Case during subsequent 120-month ISI program 
intervals will be required to implement the latest version incorporated 
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of NRC 
Regulatory Guide 1.192, as incorporated by reference in paragraph 
(a)(3)(iii) of this section.
    (iii) OM Code Case condition: Applying annulled Code Cases. 
Application of an annulled Code Case is prohibited unless a licensee 
previously applied the listed Code Case prior to it being listed as 
annulled in NRC Regulatory Guide 1.192. If a licensee has applied a 
listed Code Case that is later listed as annulled in NRC Regulatory 
Guide 1.192, the licensee may continue to apply the Code Case to the 
end of the current 120-month interval.
* * * * *
    (f) Inservice testing requirements. Systems and components of 
boiling and pressurized water-cooled nuclear power reactors must meet 
the requirements for preservice and inservice testing (referred to in 
this paragraph collectively as inservice testing) of the ASME BPV Code 
and ASME OM Code as specified in this paragraph. Each operating license 
for a boiling or pressurized water-cooled nuclear facility is subject 
to the following conditions. Each combined license for a boiling or 
pressurized water-cooled nuclear facility is subject to the following 
conditions, but the conditions in paragraphs (f)(4) through (6) of this 
section must be met only after the Commission makes the finding under 
Sec.  52.103(g) of this chapter. Requirements for inservice inspection 
of Class 1, Class 2, Class 3, Class MC, and Class CC components 
(including their supports) are located in Sec.  50.55a(g).
* * * * *
    (2) Design and accessibility requirements for performing inservice 
testing in plants with CPs issued between 1971 and 1974. For a boiling 
or pressurized water-cooled nuclear power facility whose construction 
permit was issued on or after January 1, 1971, but before July 1, 1974, 
pumps and valves that are classified as ASME Code Class 1 and Class 2 
must be designed and provided with access to enable the performance of 
inservice tests for operational readiness set forth in editions and 
addenda of Section XI of the ASME BPV incorporated by reference in 
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases 
listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 1.192, as 
incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this 
section, respectively) in effect 6 months before the date of issuance 
of the construction permit. The pumps and valves may meet the inservice 
test requirements set forth in subsequent editions of this Code and 
addenda that are incorporated by reference in paragraph (a)(1)(ii) of 
this section (or the optional ASME Code Cases listed in NRC Regulatory 
Guide 1.147 or NRC Regulatory Guide 1.192, as incorporated by reference 
in paragraphs (a)(3)(ii) and (iii) of this section, respectively), 
subject to the applicable conditions listed therein.
* * * * *
    (3) * * *
    (iii) * * *
    (A) Class 1 pumps and valves: First provision. In facilities whose 
construction permit was issued before November 22, 1999, pumps and 
valves that are classified as ASME Code Class 1 must be designed and 
provided with access to enable the performance of inservice testing of 
those pumps and valves within the scope of the ASME OM Code for 
assessing operational readiness, as set forth in either the editions 
and addenda of Section XI of the ASME BPV Code incorporated by 
reference in paragraph (a)(1)(ii) of this section (or the optional ASME 
Code Cases listed in NRC Regulatory Guide 1.147 or NRC Regulatory Guide 
1.192, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) 
of this section, respectively) which are applied to the construction of 
the particular pump or valve or the summer 1973 Addenda, whichever is 
later.
    (B) Class 1 pumps and valves: Second provision. In facilities whose 
construction permit under this part, or design certification, design 
approval, combined license, or manufacturing license under part 52 of 
this chapter, issued on or after November 22, 1999,

[[Page 56862]]

pumps and valves that are classified as ASME Code Class 1 must be 
designed and provided with access to enable the performance of 
inservice testing of those pumps and valves within the scope of the 
ASME OM Code for assessing operational readiness, as set forth in 
editions and addenda of the ASME OM Code (or the optional ASME Code 
Cases listed in NRC Regulatory Guide 1.192, as incorporated by 
reference in paragraph (a)(3)(iii) of this section), incorporated by 
reference in paragraph (a)(1)(iv) of this section at the time the 
construction permit, combined license, manufacturing license, design 
certification, or design approval is issued.
    (iv) * * *
    (A) Class 2 and 3 pumps and valves: First provision. In facilities 
whose construction permit was issued before November 22, 1999, pumps 
and valves that are classified as ASME Code Class 2 and Class 3 that 
are within the scope of the ASME OM Code and are not covered by 
paragraph (f)(3)(iii)(A) of this section must be designed and be 
provided with access to enable the performance of inservice testing of 
the pumps and valves for assessing operational readiness set forth in 
the editions and addenda of Section XI of the ASME BPV Code 
incorporated by reference in paragraph (a)(1)(ii) of this section (or 
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section) 
applied to the construction of the particular pump or valve or the 
Summer 1973 Addenda, whichever is later.
    (B) Class 2 and 3 pumps and valves: Second provision. In facilities 
whose construction permit under this part, or design certification, 
design approval, combined license, or manufacturing license under part 
52 of this chapter, issued on or after November 22, 1999, pumps and 
valves that are classified as ASME Code Class 2 and 3 that are within 
the scope of the ASME OM Code and are not covered by paragraph 
(f)(3)(iii)(B) of this section must be designed and provided with 
access to enable the performance of inservice testing of the pumps and 
valves for assessing operational readiness set forth in editions and 
addenda of the ASME OM Code (or the optional ASME OM Code Cases listed 
in NRC Regulatory Guide 1.192, as incorporated by reference in 
paragraph (a)(3)(iii) of this section), incorporated by reference in 
paragraph (a)(1)(iv) of this section at the time the construction 
permit, combined license, or design certification is issued.
* * * * *
    (4) Inservice testing standards requirement for operating plants. 
Throughout the service life of a boiling or pressurized water-cooled 
nuclear power facility, pumps and valves that are within the scope of 
the ASME OM Code must meet the inservice test requirements (except 
design and access provisions) set forth in the ASME OM Code and addenda 
that become effective subsequent to editions and addenda specified in 
paragraphs (f)(2) and (3) of this section and that are incorporated by 
reference in paragraph (a)(1)(iv) of this section, to the extent 
practical within the limitations of design, geometry, and materials of 
construction of the components.
    (i) Applicable IST Code: Initial 120-month interval. Inservice 
tests to verify operational readiness of pumps and valves, whose 
function is required for safety, conducted during the initial 120-month 
interval must comply with the requirements in the latest edition and 
addenda of the OM Code incorporated by reference in paragraph 
(a)(1)(iv) of this section on the date 12 months before the date of 
issuance of the operating license under this part, or 12 months before 
the date scheduled for initial loading of fuel under a combined license 
under part 52 of this chapter (or the optional ASME Code Cases listed 
in NRC Regulatory Guide 1.192, as incorporated by reference in 
paragraph (a)(3)(iii) of this section, subject to the conditions listed 
in paragraph (b) of this section).
    (ii) Applicable IST Code: Successive 120-month intervals. Inservice 
tests to verify operational readiness of pumps and valves, whose 
function is required for safety, conducted during successive 120-month 
intervals must comply with the requirements of the latest edition and 
addenda of the OM Code incorporated by reference in paragraph 
(a)(1)(iv) of this section 12 months before the start of the 120-month 
interval (or the optional ASME Code Cases listed in NRC Regulatory 
Guide 1.147 or NRC Regulatory Guide 1.192 as incorporated by reference 
in paragraphs (a)(3)(ii) and (iii) of this section, respectively), 
subject to the conditions listed in paragraph (b) of this section.
* * * * *
    (g) Preservice and inservice inspection requirements. Systems and 
components of boiling and pressurized water-cooled nuclear power 
reactors must meet the requirements of the ASME BPV Code as specified 
in this paragraph. Each operating license for a boiling or pressurized 
water-cooled nuclear facility is subject to the following conditions. 
Each combined license for a boiling or pressurized water-cooled nuclear 
facility is subject to the following conditions, but the conditions in 
paragraphs (g)(4) through (6) of this section must be met only after 
the Commission makes the finding under Sec.  52.103(g) of this chapter. 
Requirements for inservice testing of Class 1, Class 2, and Class 3 
pumps and valves are located in Sec.  50.55a(f).
* * * * *
    (2) Accessibility requirements--(i) Accessibility requirements for 
plants with CPs issued between 1971 and 1974. For a boiling or 
pressurized water-cooled nuclear power facility whose construction 
permit was issued on or after January 1, 1971, but before July 1, 1974, 
components that are classified as ASME Code Class 1 and Class 2 and 
supports for components that are classified as ASME Code Class 1 and 
Class 2 must be designed and be provided with the access necessary to 
perform the required preservice and inservice examinations set forth in 
editions and addenda of Section III or Section XI of the ASME BPV Code 
incorporated by reference in paragraph (a)(1) of this section (or the 
optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section) in 
effect 6 months before the date of issuance of the construction permit.
    (ii) Accessibility requirements for plants with CPs issued after 
1974. For a boiling or pressurized water-cooled nuclear power facility, 
whose construction permit under this part, or design certification, 
design approval, combined license, or manufacturing license under part 
52 of this chapter, was issued on or after July 1, 1974, components 
that are classified as ASME Code Class 1, Class 2, and Class 3 and 
supports for components that are classified as ASME Code Class 1, Class 
2, and Class 3 must be designed and provided with the access necessary 
to perform the required preservice and inservice examinations set forth 
in editions and addenda of Section III or Section XI of the ASME BPV 
Code incorporated by reference in paragraph (a)(1) of this section (or 
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section) 
applied to the construction of the particular component.
    (iii) Accessibility requirements: Meeting later Code requirements. 
All components (including supports) may meet the requirements set forth 
in subsequent editions of codes and addenda or portions thereof that 
are

[[Page 56863]]

incorporated by reference in paragraph (a) of this section, subject to 
the conditions listed therein.
    (3) Preservice examination requirements--(i) Preservice examination 
requirements for plants with CPs issued between 1971 and 1974. For a 
boiling or pressurized water-cooled nuclear power facility whose 
construction permit was issued on or after January 1, 1971, but before 
July 1, 1974, components that are classified as ASME Code Class 1 and 
Class 2 and supports for components that are classified as ASME Code 
Class 1 and Class 2 must meet the preservice examination requirements 
set forth in editions and addenda of Section III or Section XI of the 
ASME BPV Code incorporated by reference in paragraph (a)(1) of this 
section (or the optional ASME Code Cases listed in NRC Regulatory Guide 
1.147, as incorporated by reference in paragraph (a)(3)(ii) of this 
section) in effect 6 months before the date of issuance of the 
construction permit.
    (ii) Preservice examination requirements for plants with CPs issued 
after 1974. For a boiling or pressurized water-cooled nuclear power 
facility, whose construction permit under this part, or design 
certification, design approval, combined license, or manufacturing 
license under part 52 of this chapter, was issued on or after July 1, 
1974, components that are classified as ASME Code Class 1, Class 2, and 
Class 3 and supports for components that are classified as ASME Code 
Class 1, Class 2, and Class 3 must meet the preservice examination 
requirements set forth in the editions and addenda of Section III or 
Section XI of the ASME BPV Code incorporated by reference in paragraph 
(a)(1) of this section (or the optional ASME Code Cases listed in NRC 
Regulatory Guide 1.147, as incorporated by reference in paragraph 
(a)(3)(ii) of this section) applied to the construction of the 
particular component.
* * * * *
    (v) Preservice examination requirements: Meeting later Code 
requirements. All components (including supports) may meet the 
requirements set forth in subsequent editions of codes and addenda or 
portions thereof that are incorporated by reference in paragraph (a) of 
this section, subject to the conditions listed therein.
* * * * *
    (4) * * *
    (i) Applicable ISI Code: Initial 120-month interval. Inservice 
examination of components and system pressure tests conducted during 
the initial 120-month inspection interval must comply with the 
requirements in the latest edition and addenda of the Code incorporated 
by reference in paragraph (a) of this section on the date 12 months 
before the date of issuance of the operating license under this part, 
or 12 months before the date scheduled for initial loading of fuel 
under a combined license under part 52 of this chapter (or the optional 
ASME Code Cases listed in NRC Regulatory Guide 1.147, when using 
Section XI, or NRC Regulatory Guide 1.192, when using the OM Code, as 
incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this 
section, respectively), subject to the conditions listed in paragraph 
(b) of this section.
    (ii) Applicable ISI Code: Successive 120-month intervals. Inservice 
examination of components and system pressure tests conducted during 
successive 120-month inspection intervals must comply with the 
requirements of the latest edition and addenda of the Code incorporated 
by reference in paragraph (a) of this section 12 months before the 
start of the 120-month inspection interval (or the optional ASME Code 
Cases listed in NRC Regulatory Guide 1.147, when using Section XI, or 
NRC Regulatory Guide 1.192, when using the OM Code, as incorporated by 
reference in paragraphs (a)(3)(ii) and (iii) of this section), subject 
to the conditions listed in paragraph (b) of this section. However, a 
licensee whose inservice inspection interval commences during the 12 
through 18-month period after July 21, 2011, may delay the update of 
their Appendix VIII program by up to 18 months after July 21, 2011.
* * * * *
    (6) * * *
    (ii) * * *
    (D) * * *
    (1) Implementation: Holders of operating licenses or combined 
licenses for pressurized-water reactors as of or after [the effective 
date of the final rule] shall implement the requirements of ASME BPV 
Code Case N-729-4 instead of ASME BPV Code Case N-729-1, subject to the 
conditions specified in paragraphs (g)(6)(ii)(D)(2) through (4) of this 
section, by the first refueling outage starting after [the effective 
date of the final rule].
    (2) Appendix I use: Appendix I of ASME BPV Code Case N-729-4 shall 
not be implemented without prior NRC approval.
    (3) Bare metal visual frequency: Instead of Note 4 of ASME BPV Code 
Case N-729-4, the following shall be implemented; If EDY<8 and if no 
flaws are found that are attributed to PWSCC; (a) A bare metal visual 
examination is not required during refueling outages when a volumetric 
or surface examination is performed, (b) If a wetted surface 
examination has been performed of all of the partial penetration welds 
during the previous non-visual examination, the reexamination frequency 
may be extended to every third refueling outage or 5 calendar years, 
whichever is less, provided an IWA-2212 VT-2 visual examination of the 
head is performed under the insulation through multiple access points 
in outages that the VE is not completed. This IWA-2212 VT-2 visual 
examination may be performed with the reactor vessel depressurized.
    (4) Surface exam acceptance criteria: In addition to the 
requirements of paragraph-3132.1(b) of ASME BPV Code Case N-729-4, a 
component whose surface examination detects rounded indications greater 
than allowed in Paragraph NB-5352 in size on the partial-penetration or 
associated fillet weld shall be classified as having an unacceptable 
indication and corrected in accordance with the provisions of 
paragraph-3132.2 of ASME BPV Code Case N-729-4.
* * * * *
    (F) * * *
    (1) Implementation: Holders of operating licenses or combined 
licenses for pressurized-water reactors as of or after [the effective 
date of the final rule] shall implement the requirements of ASME BPV 
Code Case N-770-2 instead of ASME BPV Code Case N-770-1, subject to the 
conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of 
this section, by the first refueling outage starting after [the 
effective date of the final rule].
    (2) Categorization: Full structural weld overlays, authorized by 
the NRC staff in accordance with the alternatives approval process of 
this section, may be categorized as Inspection Items C-1 or F-1, as 
appropriate. Welds that have been mitigated by the Mechanical Stress 
Improvement Process (MSIP\TM\) may be categorized as Inspection Items D 
or E, as appropriate, provided the criteria in Appendix I of the code 
case have been met. For the purpose of determining ISI frequencies, all 
other butt welds that rely on Alloy 82/182 for structural integrity 
shall be categorized as Inspection Items A-1, A-2, or B until the NRC 
staff has reviewed the mitigation and authorized an alternative code 
case Inspection Item for the mitigated weld, or an alternative code 
case Inspection Item is used based on

[[Page 56864]]

conformance with an ASME mitigation code case endorsed in NRC 
Regulatory Guide 1.147 with any applying conditions specified in NRC 
Regulatory Guide 1.147, as incorporated by reference in paragraph 
(a)(3)(ii) of this section. Paragraph-1100(e) of ASME BPV Code Case N-
770-2 shall not be used to exempt welds that rely on Alloy 82/182 for 
structural integrity from any requirement of paragraph (g)(6)(ii)(F) of 
this section.
    (3) Baseline examinations: Baseline examinations for welds in Table 
1 of ASME BPV Code Case N-770-2, Inspection Items A-1, A-2, and B, if 
not previously performed or currently scheduled to be performed in an 
ongoing refueling outage at the time this rule becomes effective, in 
accordance with paragraph (g)(6)(ii)(F) of this section, shall be 
completed by the end of the next refueling outage. Previous 
examinations of these welds can be credited for baseline examinations 
only if they were performed within the re-inspection period for the 
weld item in Table 1 of ASME BPV Code Case N-770-2 and the examination 
of each weld meets the examination requirements of paragraphs -2500(a) 
or -2500(b) of ASME BPV Code Case N-770-2. Other previous examinations 
that do not meet these requirements can be used to meet the baseline 
examination requirement, provided NRC approval in accordance with 
paragraphs (z)(1) or (2) of this section, is granted prior to the end 
of the next refueling outage.
    (4) Examination coverage: When implementing paragraph-2500(a) of 
ASME Code Case N-770-2, essentially 100 percent volumetric examination 
coverage shall be obtained, including greater than 90 percent 
volumetric examination coverage for circumferential flaws. Licensees 
are prohibited from using Paragraph-2500(c) and -2500(d) of ASME BPV 
Code Case N-770-2 to meet examination requirements.
    (5) Inlay/onlay inspection frequency: All hot-leg operating 
temperature welds in Inspection Items G, H, J, and K shall be inspected 
each inspection interval. A 25 percent sample of Inspection Items G, H, 
J, and K cold-leg operating temperature welds shall be inspected 
whenever the core barrel is removed (unless it has already been 
inspected within the past 10 years) or within 20 years, whichever is 
less.
    (6) Reporting requirements: For any mitigated weld whose volumetric 
examination detects growth of existing flaws in the required 
examination volume that exceed the previous IWB-3600 flaw evaluations 
or new flaws, a report summarizing the evaluation, along with inputs, 
methodologies, assumptions, and causes of the new flaw or flaw growth 
is to be provided to the NRC prior to the weld being placed in service 
other than modes 5 or 6.
    (7) Defining ``t'': For Inspection Items G, H, J, and K, when 
applying the acceptance standards of ASME BPV Code, Section XI, IWB-
3514, for planar flaws contained within the inlay or onlay, the 
thickness ``t'' in IWB-3514 is the thickness of the inlay or onlay. For 
planar flaws in the balance of the dissimilar metal weld examination 
volume, the thickness ``t'' in IWB-3514 is the combined thickness of 
the inlay or onlay and the dissimilar metal weld.
    (8) Optimized weld overlay examination: Initial inservice 
examination of Inspection Item C-2 welds, shall be performed between 
the third refueling outage and no later than 10 years after application 
of the overlay.
    (9) Deferral: Note (11)(b)(1) in ASME BPV Code Case N-770-2 shall 
not be used to defer the initial inservice examination of optimized 
weld overlays (i.e., Inspection Item C-2 of ASME BPV Code Case N-770-
2).
    (10) Examination technique: Note 14(b) of Table 1 and Note (b) of 
Figure 5(a) of ASME BPV Code Case N-770-2 may only be implemented if 
the requirements of Note 14(a) of Table 1 of ASME BPV Code Case N-770-2 
cannot be met.
    (11) Cast stainless steel: Examination of ASME Code Class 1 piping 
and vessel nozzle butt welds involving cast stainless steel materials, 
shall be performed with Appendix VIII, Supplement 9 qualifications, or 
qualifications similar to Appendix VIII, Supplement 2 or 10 using cast 
stainless steel mockups no later than the next scheduled weld 
examination after January 1, 2020, in accordance with the requirements 
of paragraph-2500(a).
    (12) Stress improvement inspection coverage: Under Paragraph I.5.1, 
for cast stainless steel items, the required examination volume shall 
be examined by Appendix VIII procedures to the maximum extent practical 
including 100 percent of the susceptible material volume.
    (13) Encoded ultrasonic examination: Ultrasonic examinations 
performed in accordance with the requirements of Table 1 for Inspection 
Item A-1, A-2, B, E, F-2, J, and K shall be performed for essentially 
100 percent of the inspection surface area using an encoded method.
* * * * *

    Dated at Rockville, Maryland, this 21st day of August 2015.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2015-23193 Filed 9-17-15; 8:45 am]
 BILLING CODE 7590-01-P

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