Duke Energy Carolinas, LLC., Oconee Nuclear Station, Units 1, 2, and 3 Exemption, 26318-26321 [2012-10698]

Download as PDF 26318 Federal Register / Vol. 77, No. 86 / Thursday, May 3, 2012 / Notices With regard to potential nonradiological impacts, the proposed action does not have a potential to affect any historic sites. The proposed action does not affect non-radiological plant effluents, air quality or noise. The proposed action and attendant exemption of the material from further AEA and NRC licensing requirements will not significantly increase the probability or consequences of accidents, no changes are being made in the types of any effluents that may be released off site, and there is no significant increase in occupational or public radiation exposure. Environmental Impacts of the Alternatives to the Proposed Action Due to the very small amounts of radioactive material involved, the environmental impacts of the proposed action are small. Therefore, the only alternative the staff considered is the no-action alternative, under which the staff would deny the disposal request. This denial of the request would result in no change in current environmental impacts. The environmental impacts of the proposed action and the no-action alternative are therefore similar and the no-action alternative is accordingly not further considered. Conclusion The NRC staff has concluded that the proposed action will not significantly impact the quality of the human environment, and that the proposed action is the preferred alternative. wreier-aviles on DSK7SPTVN1PROD with NOTICES Agencies and Persons Consulted The NRC provided a draft of this Environmental Assessment to the State of Idaho Department of Environmental Quality for review on February 29, 2012. The State had no comments. The NRC staff has determined that the proposed action is of a procedural nature, and will not affect listed species or critical habitat. Therefore, no further consultation is required under Section 7 of the Endangered Species Act. The NRC staff has also determined that the proposed action is not the type of activity that has the potential to cause effects on historic properties. Therefore, no further consultation is required under Section 106 of the National Historic Preservation Act. III. Finding of No Significant Impact The NRC staff has prepared this EA in support of the proposed action. On the basis of this EA, the NRC finds that there are no significant environmental impacts from the proposed action, and that preparation of an environmental impact statement is not warranted. VerDate Mar<15>2010 15:32 May 02, 2012 Jkt 226001 Accordingly, the NRC has determined that a Finding of No Significant Impact is appropriate. IV. Further Information Documents related to this action, including the application and supporting documentation, are available online in the NRC Library at https:// www.nrc.gov/reading-rm/adams.html. From this site, you can access the NRC’s Agencywide Document Access and Management System (ADAMS), which provides text and image files of NRC’s public documents. The documents related to this action are listed below, along with their ADAMS numbers. (1) Letter dated June 7, 2011, ‘‘Humboldt Bay Power Plant Unit 3, Request for 10 CFR 20.2002 Alternate Disposal Approval and 10 CFR 30.11 Exemption of Humboldt Bay Power Plant Waste for Disposal at US Ecology Idaho [ADAMS Accession Number ML11160A211]. (2) E–Mail dated January 9, 2012, providing responses to a request for additional information and corrected information for the prior submittal [ADAMS Accession Number ML120330349]. (3) NRC letter dated November 2, 2010, approving prior request from Humboldt Bay for 10 CFR 20.2002 alternate disposal and 10 CFR 30.11 exemption [ADAMS Accession Number ML102870344]. If you do not have access to ADAMS, or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1–800–397–4209, 301– 415–4737, or by email to pdr@nrc.gov. These documents may also be viewed electronically on the public computers located at the NRC’s PDR, O 1 F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction contractor will copy documents for a fee. For the U.S. Nuclear Regulatory Commission. Dated at Rockville, Maryland, this 25th day of April, 2012. Paul Michalak, Acting Deputy Director, Decommissioning and Uranium Recovery Licensing Directorate, Division of Waste Management and Environmental Protection, Office of Federal and State Materials and Environmental Management Programs. [FR Doc. 2012–10700 Filed 5–2–12; 8:45 am] BILLING CODE 7590–01–P PO 00000 Frm 00078 Fmt 4703 Sfmt 4703 NUCLEAR REGULATORY COMMISSION [Docket Nos. 50–269, 50–270, and 50–287; NRC–2012–0088] Duke Energy Carolinas, LLC., Oconee Nuclear Station, Units 1, 2, and 3 Exemption 1.0 Background Duke Energy Carolinas, LLC (the licensee) is the holder of Renewed Facility Operating Licenses DPR–38, DPR–47, and DPR–55, which authorize operation of the Oconee Nuclear Station, Units 1, 2 and 3 (ONS, Units 1, 2, and 3). The licenses provide, among other things, that the facilities are subject to all rules, regulations, and orders of the U.S. Nuclear Regulatory Commission (NRC, the Commission) now or hereafter in effect. The facility consists of three pressurized water reactors located in Oconee County in South Carolina. 2.0 Request/Action Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix G, ‘‘Fracture Toughness Requirements,’’ requires that fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors 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; and 10 CFR 50.61, ‘‘Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,’’ provides fracture toughness requirements for protection against pressurized thermal shock (PTS) events. By letter dated August 3, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11223A010), the licensee requested exemptions from certain requirements of 10 CFR 50.61 and 10 CFR Part 50, Appendix G. The exemptions would allow use of alternate initial RTNDT (reference nil ductility temperature), as described in the NRCapproved topical reports (TRs), BAW– 2308, ‘‘Initial RTNDT of Linde 80 Weld Materials,’’ Revisions 1–A and 2–A, for determining the adjusted RTNDT of Linde 80 weld materials present in the beltline region of the ONS, Units 1, 2, and 3 reactor vessels (RVs). The licensee requested an exemption from Appendix G to 10 CFR Part 50 to replace the required use of the existing E:\FR\FM\03MYN1.SGM 03MYN1 Federal Register / Vol. 77, No. 86 / Thursday, May 3, 2012 / Notices Charpy V-notch (Cv) and drop weightbased methodology and allow the use of an alternate methodology to incorporate the use of fracture toughness test data for evaluating the integrity of the ONS, Units 1, 2, and 3 reactor vessel (RV) beltline welds based on the use of the 1997 and 2002 editions of American Society for Testing and Materials (ASTM) Standard Test Method E 1921, ‘‘Standard Test Method for Determination of Reference Temperature T0, for Ferritic Steels in the Transition Range,’’ and American Society of Mechanical Engineers (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 of Section III, Division 1, Class 1.’’ The exemption is required since Appendix G to 10 CFR Part 50, through reference to Appendix G to Section XI of the ASME Code pursuant to 10 CFR 50.55a, requires the use of a methodology based on Cv and drop weight data. The licensee also requested an exemption from 10 CFR 50.61(a)(5) to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the ONS, Units 1, 2, and 3 for RV beltline welds based on the use of the 1997 and 2002, editions of ASTM E 1921, and ASME Code Case N–629. The exemption is required since the methodology for evaluating RV material fracture toughness in 10 CFR 50.61 requires the use of the Cv and drop weight data for establishing the PTS reference temperature (RTPTS). wreier-aviles on DSK7SPTVN1PROD with NOTICES 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, are consistent with the common defense and security; and (2) when special circumstances are present. These circumstances include the special circumstances that allow 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. 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 and 10 CFR 50.61, and does not exempt the licensee from meeting any other requirement of 10 CFR Part 50, Appendix G and 10 CFR 50.61. VerDate Mar<15>2010 15:32 May 02, 2012 Jkt 226001 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 ONS, Units 1, 2, and 3 RV beltline welds, and would not result in changes to operation of the plant. Section 50.60(b) of 10 CFR Part 50 allows the use of 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. In addition, 10 CFR 50.60(b) of 10 CFR Part 50 permits different NRC approved methods for use in determining the initial material properties. As stated above, 10 CFR 50.12(a) allows the NRC to grant exemptions from 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. 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 exemptions are justified based on the licensee utilizing the fracture toughness methodology specified in BAW–2308, Revisions 1–A 1 and 2–A, which include the conditions and limitations delineated in the NRC staff’s safety evaluations (SEs), dated August 4, 2005 (ADAMS Accession No. ML052070408), and March 24, 2008 (ADAMS Accession 1 Note, a revision number including a ‘‘-A’’ denotes an NRC-staff approved version of the TR which includes the NRC staff’s final safety evaluation. PO 00000 Frm 00079 Fmt 4703 Sfmt 4703 26319 No. ML080770349). The use of the methodology specified in the NRC staff’s SEs will ensure that pressuretemperature limits developed for the ONS, Units 1, 2, and 3 RVs will continue to be based on an adequately conservative estimate of RV material properties and ensure that the pressureretaining components of the reactor coolant pressure boundary retain adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests. This exemption only modifies the methodology to be used by the licensee for demonstrating compliance with the requirements of Appendix G to 10 CFR Part 50, and does not exempt the licensee from meeting any other requirement of 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 RV materials to ensure that a licensee’s RV will be protected from failure during a PTS event. The licensee seeks an exemption from 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 TRs BAW–2308, Revisions 1–A and 2–A. These TRs established an alternative method for determining initial (unirradiated) material reference temperatures for RV 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 values may be used in lieu of the RTNDT determined as specified by paragraph NB–2331 of Section III of the ASME Code. Regulations associated with the determination of RV material properties involving protection of the RV from brittle failure or ductile rupture includes 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 E:\FR\FM\03MYN1.SGM 03MYN1 wreier-aviles on DSK7SPTVN1PROD with NOTICES 26320 Federal Register / Vol. 77, No. 86 / Thursday, May 3, 2012 / Notices determination of RTPTS, the reference temperature RTNDT, evaluated for the end of license fluence. In TR BAW–2308, Revision 1, the Babcock and Wilcox Owners Group (B&WOG) 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. 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 To values (i.e., IRTTo), 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 NRC staff’s August 4, 2005 SE of BAW–2308, Revision 1, contains the NRC staffaccepted IRTTO and initial margin (denoted as si) for specific Linde 80 weld wire heat numbers. In accordance with the conditions and limitations outlined in the NRC staff’s August 4, 2005 SE of TR BAW–2308, Revision 1, for utilizing the values in Table 3: the licensee’s proposed methodology has (1) utilized the appropriate NRC staffaccepted IRTTo and si values for Linde 80 weld wire heat numbers; (2) applied chemistry factors greater than 167 °F (the weld wire heat-specific chemical composition, via the methodology of RG 1.99, Revision 2, indicated that higher chemistry factors are applicable); (3) applied a value of 28 °F for sD in the margin term; and (4) submitted values for DRTNDT and the margin term for each Linde 80 weld in the RV 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 [reactor pressure vessels] containing weld heat 72105, and weld heat 299L44, of Linde 80 welds must use the revised IRTTo and si, values in TR BAW–2308, Revision 2.’’ The NRC staff notes that heat 299L44 is used in one ONS 1 RV beltline weld and one VerDate Mar<15>2010 15:32 May 02, 2012 Jkt 226001 ONS 2 RV beltline weld and heat. The NRC staff also notes heat 72105 is used in an ONS 3 beltline weld. The NRC staff verified that the revised IRTT0 and si values from TR BAW–2308, Revision 2 were used for these three welds. The licensee also used the revised IRTTo and si, values in TR BAW–2308, Revision 2 for the other weld heats. Although the revised IRTTo values for the weld heats other than 72105 and 299L44 are lower than the values given in the NRC staff’s SE of BAW–2308, Revision 1, these values are acceptable because the NRC staff determined in its SE for BAW– 2308, Revision 2, that the modified methodology used to calculate these values is acceptable, and more accurate than the methodology used to generate the values given in the NRC staff’s SE of BAW–2308, Revision 1. Therefore, all conditions and limitations outlined in the NRC staff SEs for TRs BAW–2308, Revisions 1 and 2, have been met for ONS, Units 1, 2, and 3. The use of the methodology in TRs BAW–2308, Revisions 1–A and 2–A, will ensure the PTS evaluation developed for the ONS, Units 1, 2, and 3 RVs will continue to be based on an adequately conservative estimate of RV material properties, and ensure the RV will be protected from failure during a PTS event. Also, when additional fracture toughness data relevant to the evaluation of the ONS, Units 1, 2, and 3 RV welds is acquired as part of the surveillance program, these data must be incorporated into the evaluation of the ONS, Units 1, 2, and 3 RV fracture toughness requirements. Based on the above, 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, the consequences of postulated accidents are not increased. Therefore, there is no undue risk to public health and safety. On February 3, 2010, a new rule, 10 CFR 50.61a, ‘‘Alternate Fracture Toughness Requirements for Protection Against [PTS] Events,’’ became effective. The NRC staff reviewed this new rule against the licensee’s exemption request and determined that there is no effect on the exemption request. The new rule does not modify the requirements from which the licensee has sought an exemption, and the alternative provided by the new rule does not address the scope of issues associated with both 10 CFR 50.61 and 10 CFR Part 50, Appendix G that the requested exemption does. PO 00000 Frm 00080 Fmt 4703 Sfmt 4703 Consistent With Common Defense and Security The proposed exemption would allow the licensee to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the ONS, Units 1, 2, and 3 RV beltline 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 reactor vessel 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 RV 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 Conclusion Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12(a), 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 Duke Energy Carolinas, LLC an exemption from certain requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61, to allow an alternative methodology to incorporate the use of fracture toughness test data for evaluating the integrity of the ONS, Units 1, 2, and 3 reactor vessel (RV) beltline welds that is based on using fracture toughness test data to determine initial, unirradiated properties. Pursuant to 10 CFR 51.32, ‘‘Finding of no significant impact,’’ the Commission has determined that the granting of this exemption will not have a significant effect on the quality of the human environment 77 FR 21594. This exemption is effective upon issuance. Dated at Rockville, Maryland, this 25th day of April 2012. E:\FR\FM\03MYN1.SGM 03MYN1 Federal Register / Vol. 77, No. 86 / Thursday, May 3, 2012 / Notices For The Nuclear Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2012–10698 Filed 5–2–12; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [Docket No. 50–288; NRC–2011–0172] Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R–112 Nuclear Regulatory Commission. ACTION: Notice of issuance. AGENCY: Please refer to Docket ID NRC–2011–0172 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly-available, using the following methods: • Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC–2011–0172. Address questions about NRC dockets to Carol Gallagher; telephone: 301–492–3668; email: Carol.Gallagher@nrc.gov. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may access publiclyavailable documents online in the NRC Library at https://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. For details with respect to the application for renewal, see the licensee’s letter dated August 29, 2007 as supplemented by letters dated January 26, July 30, October 15, 2010, and May 20, August 3, December 12, 2011, and January 27, and March 26, 2012, is available electronically under ADAMS Accession Nos. ML092310567, ML100610121, ML102360016, ML102990489, ML111520559, ML11222A026, ML113630145, ML12039A147 and ML12100A075. Also see the license’s annual reports for years 2003–2004 (ADAMS Accession No. ML043620310), 2004–2005 (ADAMS Accession No. ML052930194), 2005–2006 (ADAMS Accession No. ML062850518), 2006– 2007 (ADAMS Accession No. ML073040191), 2007–2008 (ADAMS Accession No. ML082890533), 2008– wreier-aviles on DSK7SPTVN1PROD with NOTICES ADDRESSES: VerDate Mar<15>2010 15:32 May 02, 2012 Jkt 226001 2009 (ADAMS Accession No. ML092720865), 2009–2010 (ADAMS Accession No. ML102440042), and 2010–2011 (ADAMS Accession No. ML11221A161). • 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. FOR FURTHER INFORMATION CONTACT: Geoffrey Wertz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852. Telephone: (301) 415–0893; fax number: (301) 415–3031; email: Geoffrey.Wertz@nrc.gov. SUPPLEMENTARY INFORMATION: The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued renewed Facility Operating License No. R–112, held by Reed College (the licensee), which authorizes continued operation of the Reed Research Reactor (RRR), located in Portland, Oregon. The RRR is a pool-type, natural convection, lightwater cooled, and shielded TRIGA (Training, Research, Isotope Production, General Atomics) reactor fuel. The RRR is licensed to operate at a steady-state power level of 250 kilowatts thermal power. The renewed Facility Operating License No. R–112 will expire 20 years from its date of issuance. The renewed facility operating license complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s regulations in Title 10, Chapter 1, ‘‘Nuclear Regulatory Commission,’’ of the Code of Federal Regulations (10 CFR), and sets forth those findings in the renewed facility operating license. The agency afforded an opportunity for hearing in the Notice of Opportunity for Hearing published in the Federal Register on August 19, 2011 (76 FR 52018–52022). The NRC received no request for a hearing or petition for leave to intervene following the notice. The NRC staff prepared a safety evaluation report for the renewal of Facility Operating License No. R–112 and concluded, based on that evaluation, the licensee can continue to operate the facility without endangering the health and safety of the public. The NRC staff also prepared an Environmental Assessment and Finding of No Significant Impact for the renewal of the facility operating license, noticed in the Federal Register on March 30, 2012 (77 FR 19362–19366), and concluded that renewal of the facility operating license will not have a PO 00000 Frm 00081 Fmt 4703 Sfmt 4703 26321 significant impact on the quality of the human environment. Dated at Rockville, Maryland, this 25th day of April, 2012. For the Nuclear Regulatory Commission. Jessie F. Quichocho, Chief, Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation. [FR Doc. 2012–10705 Filed 5–2–12; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [Docket Nos. 50–338 and 50–339; NRC– 2012–0051; License Nos. NPF–4 and NPF– 7] Virginia Electric and Power Company Nuclear Regulatory Commission. ACTION: Director’s Decision; issuance. AGENCY: The U.S. Nuclear Regulatory Commission (NRC or the Commission) is giving notice that the Director of the Office of Nuclear Reactor Regulation (NRR) has issued a Director’s Decision with regard to a petition dated September 8, 2011, filed by Mr. Thomas Saporito, hereinafter referred to as the ‘‘petitioner.’’ SUMMARY: Please refer to Docket ID NRC–2012–0051 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly available, using the following methods: • Federal Rulemaking Web Site: Go to https://www.regulations.gov and search for Docket ID NRC–2012–0051. Address questions about NRC dockets to Carol Gallagher; telephone: 301–492–3668; 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 https://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 notice (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 ADDRESSES: E:\FR\FM\03MYN1.SGM 03MYN1

Agencies

NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 77, Number 86 (Thursday, May 3, 2012)]
[Notices]
[Pages 26318-26321]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2012-10698]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-269, 50-270, and 50-287; NRC-2012-0088]


Duke Energy Carolinas, LLC., Oconee Nuclear Station, Units 1, 2, 
and 3 Exemption

1.0 Background

    Duke Energy Carolinas, LLC (the licensee) is the holder of Renewed 
Facility Operating Licenses DPR-38, DPR-47, and DPR-55, which authorize 
operation of the Oconee Nuclear Station, Units 1, 2 and 3 (ONS, Units 
1, 2, and 3). The licenses provide, among other things, that the 
facilities are subject to all rules, regulations, and orders of the 
U.S. Nuclear Regulatory Commission (NRC, the Commission) now or 
hereafter in effect.
    The facility consists of three pressurized water reactors located 
in Oconee County in South Carolina.

2.0 Request/Action

    Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 
Appendix G, ``Fracture Toughness Requirements,'' requires that fracture 
toughness requirements for ferritic materials of pressure-retaining 
components of the reactor coolant pressure boundary of light water 
nuclear power reactors 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; and 10 CFR 50.61, 
``Fracture Toughness Requirements for Protection Against Pressurized 
Thermal Shock Events,'' provides fracture toughness requirements for 
protection against pressurized thermal shock (PTS) events.
    By letter dated August 3, 2011 (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML11223A010), the licensee 
requested exemptions from certain requirements of 10 CFR 50.61 and 10 
CFR Part 50, Appendix G. The exemptions would allow use of alternate 
initial RTNDT (reference nil ductility temperature), as 
described in the NRC-approved topical reports (TRs), BAW-2308, 
``Initial RTNDT of Linde 80 Weld Materials,'' Revisions 1-A 
and 2-A, for determining the adjusted RTNDT of Linde 80 weld 
materials present in the beltline region of the ONS, Units 1, 2, and 3 
reactor vessels (RVs).
    The licensee requested an exemption from Appendix G to 10 CFR Part 
50 to replace the required use of the existing

[[Page 26319]]

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 ONS, 
Units 1, 2, and 3 reactor vessel (RV) beltline welds based on the use 
of the 1997 and 2002 editions of American Society for Testing and 
Materials (ASTM) Standard Test Method E 1921, ``Standard Test Method 
for Determination of Reference Temperature T0, for Ferritic 
Steels in the Transition Range,'' and American Society of Mechanical 
Engineers (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 of Section III, Division 
1, Class 1.'' The exemption is required since Appendix G to 10 CFR Part 
50, through reference to Appendix G to Section XI of the ASME Code 
pursuant to 10 CFR 50.55a, requires the use of a methodology based on 
Cv and drop weight data.
    The licensee also requested an exemption from 10 CFR 50.61(a)(5) to 
use an alternate methodology to allow the use of fracture toughness 
test data for evaluating the integrity of the ONS, Units 1, 2, and 3 
for RV beltline welds based on the use of the 1997 and 2002, editions 
of ASTM E 1921, and ASME Code Case N-629. The exemption is required 
since the methodology for evaluating RV material fracture toughness in 
10 CFR 50.61 requires the use of the Cv and drop weight data 
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, are consistent with the common defense and security; and (2) 
when special circumstances are present. These circumstances include the 
special circumstances that allow 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. 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 and 10 CFR 50.61, 
and does 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 ONS, Units 1, 2, and 3 RV beltline welds, and 
would not result in changes to operation of the plant. Section 50.60(b) 
of 10 CFR Part 50 allows the use of 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. In 
addition, 10 CFR 50.60(b) of 10 CFR Part 50 permits different NRC 
approved methods for use in determining the initial material 
properties. As stated above, 10 CFR 50.12(a) allows the NRC to grant 
exemptions from 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. 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 exemptions are justified based on the licensee 
utilizing the fracture toughness methodology specified in BAW-2308, 
Revisions 1-A \1\ and 2-A, which include the conditions and limitations 
delineated in the NRC staff's safety evaluations (SEs), dated August 4, 
2005 (ADAMS Accession No. ML052070408), and March 24, 2008 (ADAMS 
Accession No. ML080770349). The use of the methodology specified in the 
NRC staff's SEs will ensure that pressure-temperature limits developed 
for the ONS, Units 1, 2, and 3 RVs will continue to be based on an 
adequately conservative estimate of RV 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 and 
system hydrostatic tests. This exemption only modifies the methodology 
to be used by the licensee for demonstrating compliance with the 
requirements of Appendix G to 10 CFR Part 50, and does not exempt the 
licensee from meeting any other requirement of Appendix G to 10 CFR 
Part 50.
---------------------------------------------------------------------------

    \1\ Note, a revision number including a ``-A'' denotes an NRC-
staff approved version of the TR which includes the NRC staff's 
final safety evaluation.
---------------------------------------------------------------------------

    The underlying purpose of 10 CFR 50.61 is to establish requirements 
for evaluating the fracture toughness of RV materials to ensure that a 
licensee's RV will be protected from failure during a PTS event. The 
licensee seeks an exemption from 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 TRs BAW-
2308, Revisions 1-A and 2-A. These TRs established an alternative 
method for determining initial (unirradiated) material reference 
temperatures for RV 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 determined as specified by paragraph NB-2331 of 
Section III of the ASME Code. Regulations associated with the 
determination of RV material properties involving protection of the RV 
from brittle failure or ductile rupture includes 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

[[Page 26320]]

determination of RTPTS, the reference temperature 
RTNDT, evaluated for the end of license fluence.
    In TR BAW-2308, Revision 1, the Babcock and Wilcox Owners Group 
(B&WOG) 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. 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 To values (i.e., IRTTo), 
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 NRC 
staff's August 4, 2005 SE of BAW-2308, Revision 1, contains the NRC 
staff-accepted IRTTO and initial margin (denoted as 
[sigma]i) for specific Linde 80 weld wire heat numbers. In 
accordance with the conditions and limitations outlined in the NRC 
staff's August 4, 2005 SE of TR BAW-2308, Revision 1, for utilizing the 
values in Table 3: the licensee's proposed methodology has (1) utilized 
the appropriate NRC staff-accepted IRTTo and 
[sigma]i values for 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 higher chemistry factors are applicable); (3) applied 
a value of 28 [deg]F for [sigma][Delta] in the margin term; 
and (4) submitted values for [Delta]RTNDT and the margin 
term for each Linde 80 weld in the RV 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 [reactor pressure vessels] containing weld heat 
72105, and weld heat 299L44, of Linde 80 welds must use the revised 
IRTTo and [sigma]i, values in TR BAW-2308, 
Revision 2.'' The NRC staff notes that heat 299L44 is used in one ONS 1 
RV beltline weld and one ONS 2 RV beltline weld and heat. The NRC staff 
also notes heat 72105 is used in an ONS 3 beltline weld. The NRC staff 
verified that the revised IRTT0 and [sigma]i 
values from TR BAW-2308, Revision 2 were used for these three welds. 
The licensee also used the revised IRTTo and 
[sigma]i, values in TR BAW-2308, Revision 2 for the other 
weld heats. Although the revised IRTTo values for the weld 
heats other than 72105 and 299L44 are lower than the values given in 
the NRC staff's SE of BAW-2308, Revision 1, these values are acceptable 
because the NRC staff determined in its SE for BAW-2308, Revision 2, 
that the modified methodology used to calculate these values is 
acceptable, and more accurate than the methodology used to generate the 
values given in the NRC staff's SE of BAW-2308, Revision 1. Therefore, 
all conditions and limitations outlined in the NRC staff SEs for TRs 
BAW-2308, Revisions 1 and 2, have been met for ONS, Units 1, 2, and 3.
    The use of the methodology in TRs BAW-2308, Revisions 1-A and 2-A, 
will ensure the PTS evaluation developed for the ONS, Units 1, 2, and 3 
RVs will continue to be based on an adequately conservative estimate of 
RV material properties, and ensure the RV will be protected from 
failure during a PTS event. Also, when additional fracture toughness 
data relevant to the evaluation of the ONS, Units 1, 2, and 3 RV welds 
is acquired as part of the surveillance program, these data must be 
incorporated into the evaluation of the ONS, Units 1, 2, and 3 RV 
fracture toughness requirements.
    Based on the above, 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, the consequences of 
postulated accidents are not increased. Therefore, there is no undue 
risk to public health and safety. On February 3, 2010, a new rule, 10 
CFR 50.61a, ``Alternate Fracture Toughness Requirements for Protection 
Against [PTS] Events,'' became effective. The NRC staff reviewed this 
new rule against the licensee's exemption request and determined that 
there is no effect on the exemption request. The new rule does not 
modify the requirements from which the licensee has sought an 
exemption, and the alternative provided by the new rule does not 
address the scope of issues associated with both 10 CFR 50.61 and 10 
CFR Part 50, Appendix G that the requested exemption does.

Consistent With Common Defense and Security

    The proposed exemption would allow the licensee to use an alternate 
methodology to allow the use of fracture toughness test data for 
evaluating the integrity of the ONS, Units 1, 2, and 3 RV beltline 
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 reactor vessel 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 RV 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 Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), 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 Duke Energy Carolinas, LLC an 
exemption from certain requirements of Appendix G to 10 CFR Part 50 and 
10 CFR 50.61, to allow an alternative methodology to incorporate the 
use of fracture toughness test data for evaluating the integrity of the 
ONS, Units 1, 2, and 3 reactor vessel (RV) beltline welds that is based 
on using fracture toughness test data to determine initial, 
unirradiated properties.
    Pursuant to 10 CFR 51.32, ``Finding of no significant impact,'' the 
Commission has determined that the granting of this exemption will not 
have a significant effect on the quality of the human environment 77 FR 
21594.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 25th day of April 2012.

[[Page 26321]]

    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-10698 Filed 5-2-12; 8:45 am]
BILLING CODE 7590-01-P

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