Exelon Generation Company, LLC Three Mile Island Nuclear Station, Unit 1, 77726-77729 [2013-30545]

Download as PDF 77726 Federal Register / Vol. 78, No. 247 / Tuesday, December 24, 2013 / Notices good faith efforts to comply with this regulation. Because of the assumed and imposed new deadline of December 31, 2014, PPL’s exemption request seeks only temporary relief from the requirement that it file an update to the FSAR included in the BBNPP COL application. Therefore, since the relief from the requirements of 10 CFR 50.71(e)(3)(iii) would be temporary and the applicant has made good faith efforts to comply with the rule, and the underlying purpose of the rule is not served by application of the rule in this circumstance, the special circumstances required by 10 CFR 50.12(a)(2)(ii) and 10 50.12(a)(2)(v) for the granting of an exemption from 10 CFR 50.71(e)(3)(iii) exist. The proposed action involves only a schedule change which is administrative in nature. There is no consideration of any construction at this time, and hence the proposed action does not involve any construction impact. (v) There is no significant increase in the potential for or consequences from radiological accidents; and The proposed action involves only a schedule change which is administrative in nature, and does not impact the probability or consequences of accidents. (vi) The requirements from which an exemption is sought involve: (B) Reporting requirements; The exemption request involves submitting an updated FSAR by PPL and (G) Scheduling requirements; The proposed exemption relates to the schedule for submitting FSAR updates to the NRC. Eligibility for Categorical Exclusion From Environmental Review 4.0 emcdonald on DSK67QTVN1PROD with NOTICES With respect to the exemption’s impact on the quality of the human environment, the NRC has determined that this specific exemption request is eligible for categorical exclusion as identified in 10 CFR 51.22(c)(25) and justified by the NRC staff as follows: (c) The following categories of actions are categorical exclusions: (25) Granting of an exemption from the requirements of any regulation of this chapter, provided that— (i) There is no significant hazards consideration; The criteria for determining whether there is no significant hazards consideration are found in 10 CFR 50.92. The proposed action involves only a schedule change regarding the submission of an update to the application. Therefore, there is no significant hazards consideration because granting the proposed exemption would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; The proposed action involves only a schedule change which is administrative in nature and does not involve any changes to be made in the types or significant increase in the amounts of effluents that may be released offsite. (iii) There is no significant increase in individual or cumulative public or occupational radiation exposure; Since the proposed action involves only a schedule change which is administrative in nature, it does not contribute to any significant increase in occupational or public radiation exposure. (iv) There is no significant construction impact; VerDate Mar<15>2010 16:36 Dec 23, 2013 Jkt 232001 Conclusion Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12(a)(1) and (2), the exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Also special circumstances are present. Therefore, the Commission hereby grants PPL a one-time exemption from the requirements of 10 CFR 50.71(e)(3)(iii) pertaining to the Bell Bend Nuclear Power Plant COL application to allow submittal of the next FSAR update on or before December 31, 2014. Pursuant to 10 CFR 51.22, the Commission has determined that the exemption request meets the applicable categorical exclusion criteria set forth in 10 CFR 51.22(c)(25), and the granting of this exemption will not have a significant effect on the quality of the human environment. This exemption is effective upon issuance. Dated at Rockville, Maryland, this 18th day of December 2013. For The Nuclear Regulatory Commission. John Segala, Chief, Licensing Branch 1, Division of New Reactor Licensing, Office of New Reactors. [FR Doc. 2013–30752 Filed 12–23–13; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [Docket No. 50–289; NRC–2013–0274] Exelon Generation Company, LLC Three Mile Island Nuclear Station, Unit 1 Nuclear Regulatory Commission. AGENCY: PO 00000 Frm 00084 Fmt 4703 Sfmt 4703 ACTION: Exemption. Exelon Generation Company, LLC (Exelon, the licensee) is the holder of Renewed Facility Operating License No. DPR–50, which authorizes operation of the Three Mile Island Nuclear Station, Unit 1 (TMI–1). The license provides, among other things, that the facility is subject to all rules, regulations, and orders of the Nuclear Regulatory Commission (NRC) now or hereafter in effect. ADDRESSES: Please refer to Docket ID NRC–2013–0274 when contacting the NRC about the availability of information regarding this document. You may access publicly-available information related to this action by the following methods: • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC–2013–0274. Address questions about NRC dockets to Carol Gallagher; telephone: 301–287–3422; email: Carol.Gallagher@nrc.gov. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/readingrm/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. The ADAMS accession number for each document referenced in this document (if that document is available in ADAMS) is provided the first time that a document is referenced. • 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. SUPPLEMENTARY INFORMATION: SUMMARY: 1.0 Background Exelon Generation Company, LLC (Exelon, the licensee) is the holder of Renewed Facility Operating License No. DPR–50, which authorizes operation of the Three Mile Island Nuclear Station, Unit 1 (TMI–1). The license provides, among other things, that the facility is subject to all rules, regulations, and orders of the NRC now or hereafter in effect. The facility consists of a single pressurized-water reactor located in Dauphin County, Pennsylvania. 2.0 Request/Action Part 50, Appendix G of Title 10 of the Code of Federal Regulations (10 CFR), E:\FR\FM\24DEN1.SGM 24DEN1 emcdonald on DSK67QTVN1PROD with NOTICES Federal Register / Vol. 78, No. 247 / Tuesday, December 24, 2013 / Notices ‘‘Fracture Toughness Requirements,’’ specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Section 50.61, ‘‘Fracture toughness requirements for protection against pressurized thermal shock [PTS] events,’’ provides fracture toughness requirements for protection against PTS events. By letter dated December 14, 2012, (ADAMS) Accession No. ML12353A319), as supplemented by letters dated January 31, 2013, and August 13, 2013, (ADAMS Accession Nos. ML13032A312 and ML13232A214, respectively), Exelon proposed exemptions from portions of the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, to revise certain TMI–1 reactor pressure vessel (RPV) initial (unirradiated) properties using AREVA Non-Proprietary Topical Report (TR) BAW–2308, Revisions 1A and 2A, ‘‘Initial RTNDT [nil-ductility reference temperature] of Linde 80 Weld Materials.’’ The licensee requested an exemption from portions of 10 CFR Part 50, Appendix G, to replace the required use of the existing Charpy V-notch (Cv) and drop weight-based methodology and allow the use of an alternate methodology to incorporate the use of fracture toughness test data for evaluating the integrity of the TMI–1 Linde 80 weld materials in the RPV beltline. This request for exemption is based on the use of the 1997 and 2002, editions of American Society for Testing and Materials (ASTM) Standard Test Method E 1921 (ASTM E 1921), ‘‘Standard Test Method for Determination of Reference Temperature T0, for Ferritic Steels in the Transition Range,’’ and American Society for Mechanical Engineering (ASME), Boiler and Pressure Vessel Code (Code), Code Case N–629, ‘‘Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials, Section III, Division 1, Class 1.’’ Specifically, 10 CFR Part 50, Appendix G(II)(D)(i), requires that the nil-ductility reference temperature (RTNDT) be evaluated according to the procedures in the ASME Code, Section III, Division 1, ‘‘Rules for Construction of Nuclear Power Plant Components,’’ Paragraph NB–2331, ‘‘Material for Vessels.’’ These VerDate Mar<15>2010 16:36 Dec 23, 2013 Jkt 232001 procedures require the use of a methodology based on drop weight tests (NB–2331(a)(1)) and Cv test data (NB– 2331(a)(2)). In addition, 10 CFR Part 50, Appendix G,(I)(A) requires the use of methods equivalent to Appendix G to ASME Section XI, Division 1, ‘‘Rules for Inservice Inspection of Nuclear Power Plant Components,’’ which specifies the use of values that have been determined using Cv and drop weight tests described above. Therefore, an exemption from portions of 10 CFR Part 50, Appendix G, is required. The licensee also requested an exemption from portions of 10 CFR 50.61 to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the TMI–1 RPV Linde 80 beltline welds based on the use of the 1997 and 2002, editions of ASTM E 1921 and ASME Code Case N–629. Similar to the above, 10 CFR 50.61(a)(5) requires that the initial (unirradiated) RTNDT, be evaluated according to the procedures in the ASME Code, Section III, Division 1, Paragraph NB–2331. As stated previously, these procedures require the use of a methodology based on drop weight tests (NB–2331(a)(1)) and Cv test data (NB–2331(a)(2)). Therefore, the exemption is required since the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires the use of the Cv and drop weight data to determine the initial RTNDT for establishing the PTS reference temperature (RTPTS). 3.0 Discussion Pursuant to 10 CFR 50.12(a), the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR Part 50 when: (1) The exemptions are authorized by law, will not present an undue risk to public health or safety, and are consistent with the common defense and security; and (2) special circumstances are present. The special circumstance that applies to these exemptions is consistent with 10 CFR 50.12(a)(2)(ii) in that the application of the regulations in this circumstance is not necessary to achieve the underlying purpose of the rules. This special circumstance allows the licensee an exemption from the use of the Cv and drop weight-based methodology required by 10 CFR Part 50, Appendix G and 10 CFR 50.61. These exemptions only modify the methodology to be used by the licensee for demonstrating compliance with the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, and do not exempt the licensee from meeting any other requirement of PO 00000 Frm 00085 Fmt 4703 Sfmt 4703 77727 10 CFR Part 50, Appendix G and 10 CFR 50.61. Authorized by Law These exemptions would allow the licensee to use an alternate methodology to make use of fracture toughness test data for evaluating the integrity of the TMI–1 RPV Linde 80 beltline materials, and would not result in changes to operation of the plant. Section 50.60(b) allows the use of proposed alternatives to the described requirements in 10 CFR Part 50, Appendix G, or portions thereof, when an exemption is granted by the Commission under 10 CFR 50.12. As stated above, 10 CFR 50.12(a) allows the NRC to grant exemptions from portions of the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61. The NRC staff has determined that granting of the licensee’s proposed exemptions will not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission’s regulations. Therefore, the exemptions are authorized by law. No Undue Risk to Public Health and Safety The underlying purpose of Appendix G to 10 CFR Part 50 is to set forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The methodology underlying the requirements of Appendix G to 10 CFR Part 50 is based on the use of Cv and drop weight data because of reference to the ASME Code, as previously described. The licensee proposes to replace the use of the existing Cv and drop weight-based methodology by a fracture toughness-based methodology to demonstrate compliance with Appendix G to 10 CFR Part 50. The NRC staff has concluded that the requested exemption to Appendix G to 10 CFR Part 50 is justified based on the licensee utilizing the fracture toughness methodology specified in TR BAW– 2308, Revisions 1A and 2A, within the conditions and limitations delineated in the NRC staff’s safety evaluations (SEs), dated August 4, 2005, and March 24, 2008 (ADAMS Accession Nos. ML052070408 and ML080770349, respectively). The use of the methodology specified in the NRC staff’s SEs will ensure that pressuretemperature limits developed for the E:\FR\FM\24DEN1.SGM 24DEN1 emcdonald on DSK67QTVN1PROD with NOTICES 77728 Federal Register / Vol. 78, No. 247 / Tuesday, December 24, 2013 / Notices TMI–1 RPV will continue to be based on an adequately conservative estimate of RPV material properties and ensure that the pressure-retaining components of the reactor coolant pressure boundary retain adequate margins of safety during any condition of normal operation, including anticipated operational occurrences. This exemption only modifies the methodology to be used by the licensee for demonstrating compliance with the requirements of 10 CFR Part 50, Appendix G(II)(D)(i) and 10 CFR Part 50, Appendix G(I)(A), and does not exempt the licensee from meeting any other requirement of Appendix G to 10 CFR Part 50. Based on the above information, no new accident precursors are created by allowing an exemption from the use of the existing Cv and drop weight-based methodology and the use of an alternative fracture toughness-based methodology to demonstrate compliance with Appendix G to 10 CFR Part 50; thus, the probability of postulated accidents is not increased. Also, based on the above information, the consequences of postulated accidents are not increased. Therefore, there is no undue risk to public health and safety associated with the proposed exemption to Appendix G to 10 CFR Part 50. The underlying purpose of 10 CFR 50.61 is to establish requirements for evaluating the fracture toughness of RPV materials to ensure that a licensee’s RPV will be protected from failure during a PTS event. The licensee seeks an exemption from portions of 10 CFR 50.61 to use a methodology for the determination of adjusted/indexing reference temperatures. The licensee proposes to use ASME Code Case N–629 and the methodology outlined in its submittal, which are based on the use of fracture toughness data, as an alternative to the Cv and drop weightbased methodology required by 10 CFR 50.61 for establishing the initial, unirradiated properties when calculating RTPTS values. The NRC staff has concluded that the exemption is justified based on the licensee utilizing the methodology specified in the NRC staff’s SEs regarding TR BAW–2308, Revisions 1A and 2A, dated August 4, 2005, and March 24, 2008, respectively. This TR established an alternative method for determining initial (unirradiated) material reference temperatures for RPV welds manufactured using Linde 80 weld flux (i.e., ‘‘Linde 80 welds’’) and established weld wire heat-specific and Linde 80 weld generic values of this reference temperature. These weld wire heatspecific and Linde 80 weld generic VerDate Mar<15>2010 16:36 Dec 23, 2013 Jkt 232001 values may be used in lieu of the RTNDT parameter, the determination of which is specified by paragraph NB–2331 of Section III of the ASME Code. Regulations associated with the determination of RPV material properties involving protection of the RPV from brittle failure or ductile rupture include Appendix G to 10 CFR Part 50 and 10 CFR 50.61, the PTS rule. These regulations require that the initial (unirradiated) material reference temperature, RTNDT, be determined in accordance with the provisions of the ASME Code, and provide the process for determination of RTPTS, the reference temperature RTNDT, evaluated for the end of license neutron fluence. In TR BAW–2308, Revision 1, the Babcock and Wilcox Owners Group proposed to perform fracture toughness testing based on the application of the Master Curve evaluation procedure, which permits data obtained from sample sets tested at different temperatures to be combined, as the basis for redefining the initial (unirradiated) material properties of Linde 80 welds. The NRC staff evaluated this methodology for determining Linde 80 weld initial (unirradiated) material properties and uncertainty in those properties, as well as the overall method for combining unirradiated material property measurements based on T0 (initial temperature) values (i.e., initial, unirradiated nil-ductility reference temperature (IRTT0)), with property shifts from models in Regulatory Guide (RG) 1.99, Revision 2, ‘‘Radiation Embrittlement of Reactor Vessel Materials,’’ which are based on Cv testing and a defined margin term to account for uncertainties in the NRC staff SE. Table 3 in the staff’s SE of BAW–2308, Revision 1, dated August 4, 2005, contains the NRC staff-accepted IRTT0 and initial margin (denoted as si) for specific Linde 80 weld wire heat numbers. In accordance with the limitations and conditions outlined in the NRC staff’s SE of TR BAW–2308, Revision 1, dated August 4, 2005, for utilizing the values in Table 3: (1) The licensee has utilized the appropriate NRC staffaccepted IRTT0 and si values for applicable Linde 80 weld wire heat numbers; (2) applied chemistry factors greater than 167 °F (the weld wire heatspecific chemical composition, via the methodology of RG 1.99, Revision 2, indicated that chemistry factors higher than 167 °F are applicable); (3) applied a value of 28 °F for sD (i.e., shift margin) in the margin term; and (4) submitted values for DRTNDT and the margin term for each Linde 80 weld in the RPV PO 00000 Frm 00086 Fmt 4703 Sfmt 4703 through the end of the current operating license. Additionally, the NRC’s SE for TR BAW–2308, Revision 2, concludes that the revised IRTT0 and si values for Linde 80 weld materials are acceptable for referencing in plant-specific licensing applications as delineated in TR BAW–2308, Revision 2, and to the extent specified under Section 4.0, ‘‘Limitations and Conditions,’’ of the SE, which states: ‘‘Future plant-specific applications for RPVs containing weld wire heat 72105, and weld wire heat 299L44, of Linde 80 welds must use the revised IRTT0 and si, values in TR BAW–2308, Revision 2.’’ The TMI–1 RPV beltline lower nozzle belt to upper shell circumferential weld contains weld heat 72105. The following TMI–1 RPV beltline welds contain weld heat 299L44: Lower shell longitudinal weld (inner diameter 37 percent), and upper shell to lower shell circumferential weld. The licensee used the staffaccepted IRTT0 and si values for Linde 80 weld materials containing weld wire heats 299L44 and 72105. The NRC staff concludes that all conditions and limitations outlined in the NRC staff SEs for TR BAW–2308, Revisions 1A and 2A, have been met for TMI–1. The use of the methodology in TR BAW–2308, Revisions 1A and 2A, will ensure the PTS evaluation developed for the TMI–1 RPV will continue to be based on an adequately conservative estimate of RPV material properties and ensure the RPV will be protected from failure during a PTS event. The NRC staff’s SEs dated August 4, 2005, and March 24, 2008, stipulate that licensees utilize the fracture toughness methodology, specified in TR BAW– 2308, Revisions 1A and 2A, within the conditions and limitations delineated in the SEs. Based on the above information, no new accident precursors are created by allowing an exemption to use an alternate methodology to comply with the requirements of 10 CFR 50.61 in determining adjusted/indexing reference temperatures; thus, the probability of postulated accidents is not increased. Also, based on the above information, the consequences of postulated accidents are not increased. Therefore, there is no undue risk to public health and safety. Consistent With Common Defense and Security The proposed exemptions would allow the licensee to use alternate methodologies from those specified in 10 CFR Part 50, Appendix G, and 10 CFR 50.61, to allow the use of fracture toughness test data for evaluating the integrity of the TMI–1 RPV beltline E:\FR\FM\24DEN1.SGM 24DEN1 Federal Register / Vol. 78, No. 247 / Tuesday, December 24, 2013 / Notices welds. This change has no relation to security issues. Therefore, the common defense and security is not impacted by these exemptions. (3) The proposed exemption does not alter the design, function or operation of any plant equipment. Therefore, this exemption does not involve a significant reduction in a margin of safety. Special Circumstances Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii), are present whenever application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. The underlying purpose of 10 CFR Part 50, Appendix G and 10 CFR 50.61 is to protect the integrity of the reactor coolant pressure boundary by ensuring that each RPV material has adequate fracture toughness. Therefore, since the underlying purpose of 10 CFR Part 50, Appendix G and 10 CFR 50.61 is achieved by an alternative methodology for evaluating RPV material fracture toughness, the special circumstances required by 10 CFR 50(a)(2)(ii) for the granting of an exemption from portions of the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61 exist. Based on the above, the NRC has concluded that the proposed exemptions do not involve significant hazards considerations under the standards set forth in 10 CFR 50.92, and accordingly, a finding of ‘‘no significant hazards consideration’’ is justified. The NRC staff has also determined that the exemptions involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite; that there is no significant increase in individual or cumulative occupational radiation exposure; that there is no significant construction impact; and there is no significant increase in the potential for or consequences from a radiological accident. The NRC staff has further determined that the requirements from which the exemptions are sought involve the factors associated with 10 CFR 51.22(c)(25)(vi)(C)—inspection or surveillance requirements. Specifically, the exemptions address the methodology used to develop the allowable pressure and temperature criteria for determining reactor coolant system heatup/cooldown and inservice leak and hydrostatic testing in accordance with Technical Specification 3.1.2, ‘‘Pressurization Heatup and Cooldown Limitations.’’ Therefore, the criteria specified in 51.22(c)(25)(vi)(C) is satisfied and, accordingly, the exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(25). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is required to be prepared in connection with the issuance of the exemption. emcdonald on DSK67QTVN1PROD with NOTICES 4.0 Environmental Consideration The exemptions would authorize exemptions from portions of the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61 to allow the licensee to use an alternate methodology to incorporate fracture toughness test data for evaluating the integrity of the TMI–1 Linde 80 weld materials in the TMI–1 RPV beltline based on the use of the 1997 and 2002 editions of ASTM E 1921 and ASME Code Case N–629. Using the standard set forth in 10 CFR 50.92 for amendments to operating licenses, the NRC staff determined that the subject exemptions sought involve use of an alternate methodology to evaluate the integrity of the TMI–1 RPV Linde 80 beltline materials. The NRC has determined that these exemptions involve no significant hazards considerations: (1) The proposed exemptions are limited to allowing the licensee to use an alternative to the Cv and drop weight-based methodology required by 10 CFR Part 50, Appendix G and 10 CFR 50.61 to evaluate the integrity of the TMI–1 Linde 80 weld materials in the TMI– 1 RPV beltline. The alternate methodology does not involve any physical changes to the facility and does not alter the design, function or operation of any plant equipment. Therefore, issuance of this exemption does not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) The proposed exemption does not make any changes to the facility and would not create any new accident initiators. Therefore, this exemption does not create the possibility of a new or different kind of accident from any accident previously evaluated. VerDate Mar<15>2010 16:36 Dec 23, 2013 Jkt 232001 5.0 Conclusion Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12(a), the exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. Also, special circumstances are present. Therefore, the Commission hereby grants Exelon exemptions from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61, to allow an alternative methodology that is based on using fracture toughness test data to determine initial, unirradiated properties for evaluating the integrity of the TMI–1 RPV beltline welds. PO 00000 Frm 00087 Fmt 4703 Sfmt 4703 77729 This exemption is effective upon issuance. Dated at Rockville, Maryland, this 13th day of December 2013. For The Nuclear Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013–30545 Filed 12–23–13; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [NRC–2013–0273] Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations Background Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from November 28, 2013 to December 11, 2013. The last biweekly notice was published on December 10, 2013 (78 FR 74176). ADDRESSES: You may submit comment 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 http://www.regulations.gov and search for Docket ID NRC–2013–0273. Address questions about NRC dockets to Carol Gallagher; telephone: 301–287–3422; email: Carol.Gallagher@nrc.gov. • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: 3WFN, 06– 44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001. For additional direction on accessing information and submitting comments, see ‘‘Accessing Information and E:\FR\FM\24DEN1.SGM 24DEN1

Agencies

[Federal Register Volume 78, Number 247 (Tuesday, December 24, 2013)]
[Notices]
[Pages 77726-77729]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-30545]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-289; NRC-2013-0274]


Exelon Generation Company, LLC Three Mile Island Nuclear Station, 
Unit 1

AGENCY: Nuclear Regulatory Commission.

ACTION: Exemption.

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SUMMARY: Exelon Generation Company, LLC (Exelon, the licensee) is the 
holder of Renewed Facility Operating License No. DPR-50, which 
authorizes operation of the Three Mile Island Nuclear Station, Unit 1 
(TMI-1). The license provides, among other things, that the facility is 
subject to all rules, regulations, and orders of the Nuclear Regulatory 
Commission (NRC) now or hereafter in effect.

ADDRESSES: Please refer to Docket ID NRC-2013-0274 when contacting the 
NRC about the availability of information regarding this document. You 
may access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0274. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://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. The ADAMS accession number 
for each document referenced in this document (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced.
     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.

SUPPLEMENTARY INFORMATION: 

1.0 Background

    Exelon Generation Company, LLC (Exelon, the licensee) is the holder 
of Renewed Facility Operating License No. DPR-50, which authorizes 
operation of the Three Mile Island Nuclear Station, Unit 1 (TMI-1). The 
license provides, among other things, that the facility is subject to 
all rules, regulations, and orders of the NRC now or hereafter in 
effect.
    The facility consists of a single pressurized-water reactor located 
in Dauphin County, Pennsylvania.

2.0 Request/Action

    Part 50, Appendix G of Title 10 of the Code of Federal Regulations 
(10 CFR),

[[Page 77727]]

``Fracture Toughness Requirements,'' specifies fracture toughness 
requirements for ferritic materials of pressure-retaining components of 
the reactor coolant pressure boundary of light water nuclear power 
reactors to provide adequate margins of safety during any condition of 
normal operation, including anticipated operational occurrences and 
system hydrostatic tests, to which the pressure boundary may be 
subjected over its service lifetime. Section 50.61, ``Fracture 
toughness requirements for protection against pressurized thermal shock 
[PTS] events,'' provides fracture toughness requirements for protection 
against PTS events. By letter dated December 14, 2012, (ADAMS) 
Accession No. ML12353A319), as supplemented by letters dated January 
31, 2013, and August 13, 2013, (ADAMS Accession Nos. ML13032A312 and 
ML13232A214, respectively), Exelon proposed exemptions from portions of 
the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, to 
revise certain TMI-1 reactor pressure vessel (RPV) initial 
(unirradiated) properties using AREVA Non-Proprietary Topical Report 
(TR) BAW-2308, Revisions 1A and 2A, ``Initial RTNDT [nil-
ductility reference temperature] of Linde 80 Weld Materials.''
    The licensee requested an exemption from portions of 10 CFR Part 
50, Appendix G, to replace the required use of the existing Charpy V-
notch (Cv) and drop weight-based methodology and allow the 
use of an alternate methodology to incorporate the use of fracture 
toughness test data for evaluating the integrity of the TMI-1 Linde 80 
weld materials in the RPV beltline. This request for exemption is based 
on the use of the 1997 and 2002, editions of American Society for 
Testing and Materials (ASTM) Standard Test Method E 1921 (ASTM E 1921), 
``Standard Test Method for Determination of Reference Temperature 
T0, for Ferritic Steels in the Transition Range,'' and 
American Society for Mechanical Engineering (ASME), Boiler and Pressure 
Vessel Code (Code), Code Case N-629, ``Use of Fracture Toughness Test 
Data to Establish Reference Temperature for Pressure Retaining 
Materials, Section III, Division 1, Class 1.'' Specifically, 10 CFR 
Part 50, Appendix G(II)(D)(i), requires that the nil-ductility 
reference temperature (RTNDT) be evaluated according to the 
procedures in the ASME Code, Section III, Division 1, ``Rules for 
Construction of Nuclear Power Plant Components,'' Paragraph NB-2331, 
``Material for Vessels.'' These procedures require the use of a 
methodology based on drop weight tests (NB-2331(a)(1)) and 
Cv test data (NB-2331(a)(2)). In addition, 10 CFR Part 50, 
Appendix G,(I)(A) requires the use of methods equivalent to Appendix G 
to ASME Section XI, Division 1, ``Rules for Inservice Inspection of 
Nuclear Power Plant Components,'' which specifies the use of values 
that have been determined using Cv and drop weight tests 
described above. Therefore, an exemption from portions of 10 CFR Part 
50, Appendix G, is required.
    The licensee also requested an exemption from portions of 10 CFR 
50.61 to use an alternate methodology to allow the use of fracture 
toughness test data for evaluating the integrity of the TMI-1 RPV Linde 
80 beltline welds based on the use of the 1997 and 2002, editions of 
ASTM E 1921 and ASME Code Case N-629. Similar to the above, 10 CFR 
50.61(a)(5) requires that the initial (unirradiated) RTNDT, 
be evaluated according to the procedures in the ASME Code, Section III, 
Division 1, Paragraph NB-2331. As stated previously, these procedures 
require the use of a methodology based on drop weight tests (NB-
2331(a)(1)) and Cv test data (NB-2331(a)(2)). Therefore, the 
exemption is required since the methodology for evaluating RPV material 
fracture toughness in 10 CFR 50.61 requires the use of the 
Cv and drop weight data to determine the initial 
RTNDT for establishing the PTS reference temperature 
(RTPTS).

3.0 Discussion

    Pursuant to 10 CFR 50.12(a), the Commission may, upon application 
by any interested person or upon its own initiative, grant exemptions 
from the requirements of 10 CFR Part 50 when: (1) The exemptions are 
authorized by law, will not present an undue risk to public health or 
safety, and are consistent with the common defense and security; and 
(2) special circumstances are present. The special circumstance that 
applies to these exemptions is consistent with 10 CFR 50.12(a)(2)(ii) 
in that the application of the regulations in this circumstance is not 
necessary to achieve the underlying purpose of the rules. This special 
circumstance allows the licensee an exemption from the use of the 
Cv and drop weight-based methodology required by 10 CFR Part 
50, Appendix G and 10 CFR 50.61. These exemptions only modify the 
methodology to be used by the licensee for demonstrating compliance 
with the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, 
and do not exempt the licensee from meeting any other requirement of 10 
CFR Part 50, Appendix G and 10 CFR 50.61.

Authorized by Law

    These exemptions would allow the licensee to use an alternate 
methodology to make use of fracture toughness test data for evaluating 
the integrity of the TMI-1 RPV Linde 80 beltline materials, and would 
not result in changes to operation of the plant. Section 50.60(b) 
allows the use of proposed alternatives to the described requirements 
in 10 CFR Part 50, Appendix G, or portions thereof, when an exemption 
is granted by the Commission under 10 CFR 50.12. As stated above, 10 
CFR 50.12(a) allows the NRC to grant exemptions from portions of the 
requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61. The NRC 
staff has determined that granting of the licensee's proposed 
exemptions will not result in a violation of the Atomic Energy Act of 
1954, as amended, or the Commission's regulations. Therefore, the 
exemptions are authorized by law.

No Undue Risk to Public Health and Safety

    The underlying purpose of Appendix G to 10 CFR Part 50 is to set 
forth fracture toughness requirements for ferritic materials of 
pressure-retaining components of the reactor coolant pressure boundary 
of light water nuclear power reactors to provide adequate margins of 
safety during any condition of normal operation, including anticipated 
operational occurrences and system hydrostatic tests, to which the 
pressure boundary may be subjected over its service lifetime. The 
methodology underlying the requirements of Appendix G to 10 CFR Part 50 
is based on the use of Cv and drop weight data because of 
reference to the ASME Code, as previously described. The licensee 
proposes to replace the use of the existing Cv and drop 
weight-based methodology by a fracture toughness-based methodology to 
demonstrate compliance with Appendix G to 10 CFR Part 50.
    The NRC staff has concluded that the requested exemption to 
Appendix G to 10 CFR Part 50 is justified based on the licensee 
utilizing the fracture toughness methodology specified in TR BAW-2308, 
Revisions 1A and 2A, within the conditions and limitations delineated 
in the NRC staff's safety evaluations (SEs), dated August 4, 2005, and 
March 24, 2008 (ADAMS Accession Nos. ML052070408 and ML080770349, 
respectively). The use of the methodology specified in the NRC staff's 
SEs will ensure that pressure-temperature limits developed for the

[[Page 77728]]

TMI-1 RPV will continue to be based on an adequately conservative 
estimate of RPV material properties and ensure that the pressure-
retaining components of the reactor coolant pressure boundary retain 
adequate margins of safety during any condition of normal operation, 
including anticipated operational occurrences. This exemption only 
modifies the methodology to be used by the licensee for demonstrating 
compliance with the requirements of 10 CFR Part 50, Appendix 
G(II)(D)(i) and 10 CFR Part 50, Appendix G(I)(A), and does not exempt 
the licensee from meeting any other requirement of Appendix G to 10 CFR 
Part 50.
    Based on the above information, no new accident precursors are 
created by allowing an exemption from the use of the existing 
Cv and drop weight-based methodology and the use of an 
alternative fracture toughness-based methodology to demonstrate 
compliance with Appendix G to 10 CFR Part 50; thus, the probability of 
postulated accidents is not increased. Also, based on the above 
information, the consequences of postulated accidents are not 
increased. Therefore, there is no undue risk to public health and 
safety associated with the proposed exemption to Appendix G to 10 CFR 
Part 50.
    The underlying purpose of 10 CFR 50.61 is to establish requirements 
for evaluating the fracture toughness of RPV materials to ensure that a 
licensee's RPV will be protected from failure during a PTS event. The 
licensee seeks an exemption from portions of 10 CFR 50.61 to use a 
methodology for the determination of adjusted/indexing reference 
temperatures. The licensee proposes to use ASME Code Case N-629 and the 
methodology outlined in its submittal, which are based on the use of 
fracture toughness data, as an alternative to the Cv and 
drop weight-based methodology required by 10 CFR 50.61 for establishing 
the initial, unirradiated properties when calculating RTPTS 
values. The NRC staff has concluded that the exemption is justified 
based on the licensee utilizing the methodology specified in the NRC 
staff's SEs regarding TR BAW-2308, Revisions 1A and 2A, dated August 4, 
2005, and March 24, 2008, respectively. This TR established an 
alternative method for determining initial (unirradiated) material 
reference temperatures for RPV welds manufactured using Linde 80 weld 
flux (i.e., ``Linde 80 welds'') and established weld wire heat-specific 
and Linde 80 weld generic values of this reference temperature. These 
weld wire heat-specific and Linde 80 weld generic values may be used in 
lieu of the RTNDT parameter, the determination of which is 
specified by paragraph NB-2331 of Section III of the ASME Code. 
Regulations associated with the determination of RPV material 
properties involving protection of the RPV from brittle failure or 
ductile rupture include Appendix G to 10 CFR Part 50 and 10 CFR 50.61, 
the PTS rule. These regulations require that the initial (unirradiated) 
material reference temperature, RTNDT, be determined in 
accordance with the provisions of the ASME Code, and provide the 
process for determination of RTPTS, the reference 
temperature RTNDT, evaluated for the end of license neutron 
fluence.
    In TR BAW-2308, Revision 1, the Babcock and Wilcox Owners Group 
proposed to perform fracture toughness testing based on the application 
of the Master Curve evaluation procedure, which permits data obtained 
from sample sets tested at different temperatures to be combined, as 
the basis for redefining the initial (unirradiated) material properties 
of Linde 80 welds. The NRC staff evaluated this methodology for 
determining Linde 80 weld initial (unirradiated) material properties 
and uncertainty in those properties, as well as the overall method for 
combining unirradiated material property measurements based on 
T0 (initial temperature) values (i.e., initial, unirradiated 
nil-ductility reference temperature (IRTT0)), with property 
shifts from models in Regulatory Guide (RG) 1.99, Revision 2, 
``Radiation Embrittlement of Reactor Vessel Materials,'' which are 
based on Cv testing and a defined margin term to account for 
uncertainties in the NRC staff SE. Table 3 in the staff's SE of BAW-
2308, Revision 1, dated August 4, 2005, contains the NRC staff-accepted 
IRTT0 and initial margin (denoted as [sigma]i) 
for specific Linde 80 weld wire heat numbers.
    In accordance with the limitations and conditions outlined in the 
NRC staff's SE of TR BAW-2308, Revision 1, dated August 4, 2005, for 
utilizing the values in Table 3: (1) The licensee has utilized the 
appropriate NRC staff-accepted IRTT0 and [sigma]i 
values for applicable Linde 80 weld wire heat numbers; (2) applied 
chemistry factors greater than 167 [deg]F (the weld wire heat-specific 
chemical composition, via the methodology of RG 1.99, Revision 2, 
indicated that chemistry factors higher than 167 [deg]F are 
applicable); (3) applied a value of 28 [deg]F for 
[sigma][Delta] (i.e., shift margin) in the margin term; and 
(4) submitted values for [Delta]RTNDT and the margin term 
for each Linde 80 weld in the RPV through the end of the current 
operating license. Additionally, the NRC's SE for TR BAW-2308, Revision 
2, concludes that the revised IRTT0 and [sigma]i 
values for Linde 80 weld materials are acceptable for referencing in 
plant-specific licensing applications as delineated in TR BAW-2308, 
Revision 2, and to the extent specified under Section 4.0, 
``Limitations and Conditions,'' of the SE, which states: ``Future 
plant-specific applications for RPVs containing weld wire heat 72105, 
and weld wire heat 299L44, of Linde 80 welds must use the revised 
IRTT0 and [sigma]i, values in TR BAW-2308, 
Revision 2.'' The TMI-1 RPV beltline lower nozzle belt to upper shell 
circumferential weld contains weld heat 72105. The following TMI-1 RPV 
beltline welds contain weld heat 299L44: Lower shell longitudinal weld 
(inner diameter 37 percent), and upper shell to lower shell 
circumferential weld. The licensee used the staff-accepted 
IRTT0 and [sigma]i values for Linde 80 weld 
materials containing weld wire heats 299L44 and 72105. The NRC staff 
concludes that all conditions and limitations outlined in the NRC staff 
SEs for TR BAW-2308, Revisions 1A and 2A, have been met for TMI-1.
    The use of the methodology in TR BAW-2308, Revisions 1A and 2A, 
will ensure the PTS evaluation developed for the TMI-1 RPV will 
continue to be based on an adequately conservative estimate of RPV 
material properties and ensure the RPV will be protected from failure 
during a PTS event. The NRC staff's SEs dated August 4, 2005, and March 
24, 2008, stipulate that licensees utilize the fracture toughness 
methodology, specified in TR BAW-2308, Revisions 1A and 2A, within the 
conditions and limitations delineated in the SEs.
    Based on the above information, no new accident precursors are 
created by allowing an exemption to use an alternate methodology to 
comply with the requirements of 10 CFR 50.61 in determining adjusted/
indexing reference temperatures; thus, the probability of postulated 
accidents is not increased. Also, based on the above information, the 
consequences of postulated accidents are not increased. Therefore, 
there is no undue risk to public health and safety.

Consistent With Common Defense and Security

    The proposed exemptions would allow the licensee to use alternate 
methodologies from those specified in 10 CFR Part 50, Appendix G, and 
10 CFR 50.61, to allow the use of fracture toughness test data for 
evaluating the integrity of the TMI-1 RPV beltline

[[Page 77729]]

welds. This change has no relation to security issues. Therefore, the 
common defense and security is not impacted by these exemptions.

Special Circumstances

    Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii), 
are present whenever application of the regulation in the particular 
circumstances is not necessary to achieve the underlying purpose of the 
rule. The underlying purpose of 10 CFR Part 50, Appendix G and 10 CFR 
50.61 is to protect the integrity of the reactor coolant pressure 
boundary by ensuring that each RPV material has adequate fracture 
toughness. Therefore, since the underlying purpose of 10 CFR Part 50, 
Appendix G and 10 CFR 50.61 is achieved by an alternative methodology 
for evaluating RPV material fracture toughness, the special 
circumstances required by 10 CFR 50(a)(2)(ii) for the granting of an 
exemption from portions of the requirements of 10 CFR Part 50, Appendix 
G and 10 CFR 50.61 exist.

4.0 Environmental Consideration

    The exemptions would authorize exemptions from portions of the 
requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61 to allow 
the licensee to use an alternate methodology to incorporate fracture 
toughness test data for evaluating the integrity of the TMI-1 Linde 80 
weld materials in the TMI-1 RPV beltline based on the use of the 1997 
and 2002 editions of ASTM E 1921 and ASME Code Case N-629. Using the 
standard set forth in 10 CFR 50.92 for amendments to operating 
licenses, the NRC staff determined that the subject exemptions sought 
involve use of an alternate methodology to evaluate the integrity of 
the TMI-1 RPV Linde 80 beltline materials. The NRC has determined that 
these exemptions involve no significant hazards considerations:

    (1) The proposed exemptions are limited to allowing the licensee 
to use an alternative to the Cv and drop weight-based 
methodology required by 10 CFR Part 50, Appendix G and 10 CFR 50.61 
to evaluate the integrity of the TMI-1 Linde 80 weld materials in 
the TMI-1 RPV beltline. The alternate methodology does not involve 
any physical changes to the facility and does not alter the design, 
function or operation of any plant equipment. Therefore, issuance of 
this exemption does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) The proposed exemption does not make any changes to the 
facility and would not create any new accident initiators. 
Therefore, this exemption does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (3) The proposed exemption does not alter the design, function 
or operation of any plant equipment. Therefore, this exemption does 
not involve a significant reduction in a margin of safety.

    Based on the above, the NRC has concluded that the proposed 
exemptions do not involve significant hazards considerations under the 
standards set forth in 10 CFR 50.92, and accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    The NRC staff has also determined that the exemptions involve no 
significant increase in the amounts, and no significant change in the 
types, of any effluents that may be released offsite; that there is no 
significant increase in individual or cumulative occupational radiation 
exposure; that there is no significant construction impact; and there 
is no significant increase in the potential for or consequences from a 
radiological accident.
    The NRC staff has further determined that the requirements from 
which the exemptions are sought involve the factors associated with 10 
CFR 51.22(c)(25)(vi)(C)--inspection or surveillance requirements. 
Specifically, the exemptions address the methodology used to develop 
the allowable pressure and temperature criteria for determining reactor 
coolant system heatup/cooldown and inservice leak and hydrostatic 
testing in accordance with Technical Specification 3.1.2, 
``Pressurization Heatup and Cooldown Limitations.'' Therefore, the 
criteria specified in 51.22(c)(25)(vi)(C) is satisfied and, 
accordingly, the exemption meets the eligibility criteria for 
categorical exclusion set forth in 10 CFR 51.22(c)(25). Pursuant to 10 
CFR 51.22(b), no environmental impact statement or environmental 
assessment is required to be prepared in connection with the issuance 
of the exemption.

5.0 Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the exemptions are authorized by law, will not present an 
undue risk to the public health and safety, and are consistent with the 
common defense and security. Also, special circumstances are present. 
Therefore, the Commission hereby grants Exelon exemptions from the 
requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61, to allow 
an alternative methodology that is based on using fracture toughness 
test data to determine initial, unirradiated properties for evaluating 
the integrity of the TMI-1 RPV beltline welds.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 13th day of December 2013.

    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-30545 Filed 12-23-13; 8:45 am]
BILLING CODE 7590-01-P