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| {{#Wiki_filter:UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 December 22, 2011 Mr. Joseph Corporate Manager, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SEOUOYAH NUCLEAR PLANT, UNIT 2 -SAFETY EVALUATION OF RELIEF REOUEST NO. 2-APPJ-1, PROPOSED ALTERNATIVE TO THE LEAKAGE TESTING REOUIREMENTS IN PARAGRAPH IWE-S221 OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, 2001 EDITION WITH 2003 ADDENDA (TAC NO. MES24S) | | {{#Wiki_filter:UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 December 22, 2011 Mr. Joseph W. Shea Corporate Manager, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SEOUOYAH NUCLEAR PLANT, UNIT 2 -SAFETY EVALUATION OF RELIEF REOUEST NO. 2-APPJ-1, PROPOSED ALTERNATIVE TO THE LEAKAGE TESTING REOUIREMENTS IN PARAGRAPH IWE-S221 OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, 2001 EDITION WITH 2003 ADDENDA (TAC NO. MES24S) Dear Mr. Shea: By letter dated December 22,2010 (Agencywide Documents Access and Management System Accession No. ML 103630044) Tennessee Valley Authority (TVA), the licensee, submitted Relief Request 2-APPJ-1 for the Sequoyah Nuclear Plant (SON), Unit 2. Pursuant to Title 10 of the Code ofFederal Regulations (10 CFR) SO.SSa(a)(3)(i), TVA requested to utilize an alternative to the testing requirements in paragraph IWE-S221, Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2001 edition with the 2003 addenda. TVA's request for relief proposes alternative test methods (pneumatic leakage test) for containment vessel repair activities following the SON, Unit 2 Cycle 18 steam generator replacement refueling outage in fall 2012. The proposed alternative to the requirements of paragraph IWE-S221 will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR SO.SSa(a)(3(i>>, the Nuclear Regulatory Commission authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2. |
| | J. W. Shea All other requirements of the ASME Code, Sections III and XI, for which relief has not been specifically requested and approved remain applicable. If you have any questions, please feel free to contact the SON Project Manager, Siva P. Lingam, at (301) 415-1564. Sincerely,/4;:;70/.. .. /J ) -Ii*****--** (crY Douglas A. Broaddus, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328 Enclosure: Safety Evaluation cc wfencl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 2-APPJ-1 ALTERNATIVE TEST METHODS FOR CONTAINMENT REPAIR ACTIVITIES TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 DOCKET NO. 50-328 1.0 INTRODUCTION By the letter dated December 22, 2010 (Agencywide Documents Access and Management System Accession No. ML103630044), Tennessee Valley Authority (TVA, the licensee) proposed an alternative, which would allow the licensee to perform a local leak rate test of the containment in lieu of the "Leakage Test" requirements in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Code (ASME Code). The leakage tests are required as a result of planned repairs to the Sequoyah Nuclear Plant (SQN), Unit 2 steel containment vessel (SCV) in accordance with Title 10 of the Code ofFederal Regulations (10 CFR), 50.55a(a)(3)(i). The licensee will be replacing the SQN, Unit 2 steam generators during the cycle 18 refueling outage, commencing in the fall 2012. This safety evaluation addresses the ability of the proposed alternative to ensure the continued ability of the SCV to provide an acceptable level of quality and safety after the steam generator replacement activity. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested whether they could use an alternative to the requirements in Section XI of the ASME Code for post repair leakage inspection of the SQN, Unit 2 SCV from the scheduled steam generator replacement project scheduled to start in October 2012. The regulations at 10 CFR 50.55a, "Codes and Standards," incorporate, by reference, the 2001 edition through the 2003 addenda of Section XI of the ASME Code. Paragraph IWE-5221 of Subsection IWE of the ASME Code, Section XI, requires a leakage rate test following any repair and replacement activity. Paragraph IWE-5221 specifies that the leakage rate test be conducted in accordance with the provisions of 10 CFR Part 50, Appendix J, paragraph IV.A, "Containment modification," which states in part, "Any major modification, replacement of a component, which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable for the area affected by the modification," |
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| ==Dear Mr. Shea:== | | ==3.0 TECHNICAL EVALUATION== |
| By letter dated December 22,2010 (Agencywide Documents Access and Management System Accession No. ML 103630044) Tennessee Valley Authority (TVA), the licensee, submitted Relief Request 2-APPJ-1 for the Sequoyah Nuclear Plant (SON), Unit 2. Pursuant to Title 10 of the Code ofFederal Regulations (10 CFR) SO.SSa(a)(3)(i), TVA requested to utilize an alternative to the testing requirements in paragraph IWE-S221, Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2001 edition with the 2003 addenda. TVA's request for relief proposes alternative test methods (pneumatic leakage test) for containment vessel repair activities following the SON, Unit 2 Cycle 18 steam generator replacement refueling outage in fall 2012. The proposed alternative to the requirements of paragraph IWE-S221 will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR SO.SSa(a)(3(i>>, the Nuclear Regulatory Commission authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2.
| | 3.1 Specific Alternative Request The licensee proposes to perform an "as-left" local leak test on the new pressure boundary welds of the SCV repair welds in lieu of the Type A, Type B, or Type C leakage rate tests as specified in Paragraph IWE-5221 of the ASME Code for this type of repair activity. 3.2 Proposed Alternative Duration The duration of the proposed alternative is through the completion and approval of all testing associated with the restoration of the containment opening created to support the SON, Unit 2 steam generator replacement. 3.3 Basis for Relief The licensee committed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of Section III, ofthe ASME Code, 1968 Edition through the winter 1968 addenda. Before performing the repair weld, the surfaces to be welded will be cleaned. A magnetic particle test of the back gouge of the root pass will be performed, and 100 percent radiography will be performed on the pressure boundary of SCV final repair welds. In addition, ASME Code, Section XI, as modified by 10 CFR 50.55a(b)(2)(ix)(G), requires both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12.0 pounds per square inch gauge (psig) for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated. Based on the above procedure, the licensee stated that any local leakage will be directly identified by the bubble test and it will confirm the findings of the prior nondestructive examination tests. The combination of the 100 percent radiography, demonstrating that the repair welds meet the construction code radiography acceptance criteria; and the local leak rate test of the repair welds, by performing the bubble test while the SCV and repair welds are at accident pressure, will provide a more effective examination than the Type A test, as required by the ASME Code, paragraph IWE-S221. Therefore, an alternative to the requirement of paragraph IWE-S221 is requested pursuant to 10 CFR SO.55a(a)(3)(i), in that the proposed alternative provides an acceptable level of quality and safety. 3.4 Nuclear Regulatory Commission (NRC) Staff Evaluation To facilitate the SON, Unit 2 steam generator replacement, the SCV will be breached. An opening will be cut in the SCV in order to remove and replace the steam generators. After the steam generator replacement, the SCV sections removed will be reattached through welding. Paragraph IWE-5221 of the ASME Code requires that leakage rate testing be conducted to ensure the integrity of the repairs before returning the SCV to operable status. In lieu of the Type A, Type B, or Type C leakage rate testing, the licensee proposed to perform a series of examinations and a leak test subjecting the SCV to accident pressure, to verify the leak tightness and integrity of the liner welds and the SCV. |
| | | -The licensee's proposed relief request includes the detailed examination and test sequence that is summarized herein. The licensee proposed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of the ASME Code, Section III, 1968 edition through the winter 1968 addenda (TVA's code of record requirements). Before performing the repair weld, the surfaces to be welded will be cleaned and examined by magnetic particle testing of the weld preparation area, and 100 percent radiography of the final repair weld will be performed. In addition, both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds will be conducted. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12 psig for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated The magnetic particle or liquid penetrate testing, the 100 percent radiography of the repair weld, and the subsequent bubble test will provide adequate assurance that the repair welds do not leak or have any structural defects. The zero leakage acceptance criterion for the bubble test will ensure that the SCV leakage rate is not altered by the steam generator replacement activity, and the pressurization of the SCV to the accident pressure will confirm the integrity of the SCV after the repair. Therefore, the NRC staff concludes that the proposed alternative will provide adequate assurance of structural integrity. 4.0 CONCLUSION On the basis of the information provided in the relief request, and an independent NRC staff evaluation, the NRC staff concludes that the proposed alternative of a local leak rate test in lieu of the requirements in paragraph IWE-5221 of the ASME Code will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3(i)), the NRC authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2. All other requirements of the ASME Code, Section III and Section XI, for which relief has not been specifically requested and approved, remain applicable. Principal Contributor: D. Hoang Date: December 22,2011 J. W. Shea -2 All other requirements of the ASME Code, Sections III and XI, for which relief has not been specifically requested and approved remain applicable. If you have any questions, please feel free to contact the SON Project Manager, Siva P. Lingam, at (301) 415-1564. Sincerely, IRA by TOrfforl Douglas A. Broaddus, Chief Plant licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Distribution: PUBLIC LPL2-2 r/f RidsOgcMailCenter RidsNrrDorlDpr RidsAcrsAcnw MailCTR RidsNrrDorlLPL2-2 RidsNrrLACSola RidsNrrPMSequoyah RidsRgn2MailCenter RidsNrrDeEmcb RidsNrrDssScvb D. Hoang, NRR ADAMS Accession No.: ML11313A165 *by memo dated OFFICE LPL2-2/PM LPL2-2/LA EMCB/BC SCVB/BC NAME SLingam RSoia MKhanna* RDennig DBroaddus DATE 11/29/11 11/29/11 10/26/11* 12/7/11 12/22/11 OFFICIAL RECORD COPY |
| J. All other requirements of the ASME Code, Sections III and XI, for which relief has not been specifically requested and approved remain applicable. If you have any questions, please feel free to contact the SON Project Manager, Siva P. Lingam, at (301) 415-1564.
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| Sincerely,/4;:;70/.. .. /J ) -Ii*****--** (crY Douglas A. Broaddus, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328
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| ===Enclosure:===
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| Safety Evaluation cc wfencl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 2-APPJ-1 ALTERNATIVE TEST METHODS FOR CONTAINMENT REPAIR ACTIVITIES TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 DOCKET NO. 50-328 1.0 INTRODUCTION By the letter dated December 22, 2010 (Agencywide Documents Access and Management System Accession No. ML103630044), Tennessee Valley Authority (TVA, the licensee) proposed an alternative, which would allow the licensee to perform a local leak rate test of the containment in lieu of the "Leakage Test" requirements in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Code (ASME Code). The leakage tests are required as a result of planned repairs to the Sequoyah Nuclear Plant (SQN), Unit 2 steel containment vessel (SCV) in accordance with Title 10 of the Code ofFederal Regulations (10 CFR), 50.55a(a)(3)(i). The licensee will be replacing the SQN, Unit 2 steam generators during the cycle 18 refueling outage, commencing in the fall 2012. This safety evaluation addresses the ability of the proposed alternative to ensure the continued ability of the SCV to provide an acceptable level of quality and safety after the steam generator replacement activity. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested whether they could use an alternative to the requirements in Section XI of the ASME Code for post repair leakage inspection of the SQN, Unit 2 SCV from the scheduled steam generator replacement project scheduled to start in October 2012. The regulations at 10 CFR 50.55a, "Codes and Standards," incorporate, by reference, the 2001 edition through the 2003 addenda of Section XI of the ASME Code. Paragraph IWE-5221 of Subsection IWE of the ASME Code, Section XI, requires a leakage rate test following any repair and replacement activity. Paragraph IWE-5221 specifies that the leakage rate test be conducted in accordance with the provisions of 10 CFR Part 50, Appendix J, paragraph IV.A, "Containment modification," which states in part, "Any major modification, replacement of a component, which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable for the area affected by the modification,"
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| 3.0 TECHNICAL EVALUATION 3.1 Specific Alternative Request The licensee proposes to perform an "as-left" local leak test on the new pressure boundary welds of the SCV repair welds in lieu of the Type A, Type B, or Type C leakage rate tests as specified in Paragraph IWE-5221 of the ASME Code for this type of repair activity. 3.2 Proposed Alternative Duration The duration of the proposed alternative is through the completion and approval of all testing associated with the restoration of the containment opening created to support the SON, Unit 2 steam generator replacement. 3.3 Basis for Relief The licensee committed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of Section III, ofthe ASME Code, 1968 Edition through the winter 1968 addenda. Before performing the repair weld, the surfaces to be welded will be cleaned. A magnetic particle test of the back gouge of the root pass will be performed, and 100 percent radiography will be performed on the pressure boundary of SCV final repair welds. In addition, ASME Code, Section XI, as modified by 10 CFR 50.55a(b)(2)(ix)(G), requires both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12.0 pounds per square inch gauge (psig) for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated. Based on the above procedure, the licensee stated that any local leakage will be directly identified by the bubble test and it will confirm the findings of the prior nondestructive examination tests. The combination of the 100 percent radiography, demonstrating that the repair welds meet the construction code radiography acceptance criteria; and the local leak rate test of the repair welds, by performing the bubble test while the SCV and repair welds are at accident pressure, will provide a more effective examination than the Type A test, as required by the ASME Code, paragraph IWE-S221. Therefore, an alternative to the requirement of paragraph IWE-S221 is requested pursuant to 10 CFR SO.55a(a)(3)(i), in that the proposed alternative provides an acceptable level of quality and safety. 3.4 Nuclear Regulatory Commission (NRC) Staff Evaluation To facilitate the SON, Unit 2 steam generator replacement, the SCV will be breached. An opening will be cut in the SCV in order to remove and replace the steam generators. After the steam generator replacement, the SCV sections removed will be reattached through welding. Paragraph IWE-5221 of the ASME Code requires that leakage rate testing be conducted to ensure the integrity of the repairs before returning the SCV to operable status. In lieu of the Type A, Type B, or Type C leakage rate testing, the licensee proposed to perform a series of examinations and a leak test subjecting the SCV to accident pressure, to verify the leak tightness and integrity of the liner welds and the SCV.
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| -The licensee's proposed relief request includes the detailed examination and test sequence that is summarized herein. The licensee proposed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of the ASME Code, Section III, 1968 edition through the winter 1968 addenda (TVA's code of record requirements). Before performing the repair weld, the surfaces to be welded will be cleaned and examined by magnetic particle testing of the weld preparation area, and 100 percent radiography of the final repair weld will be performed. In addition, both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds will be conducted. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12 psig for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated The magnetic particle or liquid penetrate testing, the 100 percent radiography of the repair weld, and the subsequent bubble test will provide adequate assurance that the repair welds do not leak or have any structural defects. The zero leakage acceptance criterion for the bubble test will ensure that the SCV leakage rate is not altered by the steam generator replacement activity, and the pressurization of the SCV to the accident pressure will confirm the integrity of the SCV after the repair. Therefore, the NRC staff concludes that the proposed alternative will provide adequate assurance of structural integrity. 4.0 CONCLUSION On the basis of the information provided in the relief request, and an independent NRC staff evaluation, the NRC staff concludes that the proposed alternative of a local leak rate test in lieu of the requirements in paragraph IWE-5221 of the ASME Code will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3(i)), the NRC authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2. All other requirements of the ASME Code, Section III and Section XI, for which relief has not been specifically requested and approved, remain applicable. Principal Contributor: D. Hoang Date: December 22,2011}}
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MONTHYEARML1036300442010-12-22022 December 2010 Request to Employ Alternative Testing to IWE-5221 Requirements in American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI (2001 Edition with 2003 Addenda) - Request Number 2-APPJ-1 Project stage: Request ML1101900292011-01-19019 January 2011 Acceptance Review for Sequoyah-2 RR Project stage: Acceptance Review ML11313A1652011-12-22022 December 2011 Safety Evaluation of Relief Request No. 2-APP-J, Proposed Alternative to the Leakage Testing Requirements in Paragraph IWE-5221 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI Project stage: Approval 2011-01-19
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Category:Letter
MONTHYEARML24032A0202024-01-31031 January 2024 NPDES Biocide/Corrosion Treatment Plan Annual Report, Cy 2023 ML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions ML24018A0142024-01-17017 January 2024 Engine Systems, Inc., Report No. 10CFR21-0137, Rev. 1, 56913-EN 56913 ML24011A3182024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), October 2023 ML24011A3172024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), September 2023 ML24011A3202024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), December 2023 ML24011A3162024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), August 2023 ML24011A3192024-01-11011 January 2024 Submittal of Discharge Monitoring Report (Dmr), November 2023 IR 05000327/20234422024-01-11011 January 2024 95001 Supplemental Inspection Report 05000327/2023442 and 05000328/2023442 and Follow-Up Assessment Letter ML24010A2132024-01-10010 January 2024 CFR 21.21 Final Report Regarding Siemens Medium Voltage Circuit Breakers ML24018A0952024-01-0404 January 2024 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance Report No. 10CFR21-0137, Rev. 0 ML24004A0332024-01-0303 January 2024 Interim Report of a Deviation or Failure to Comply Crompton Instruments Type 077 Ammeter ML24004A0402024-01-0303 January 2024 Response to NRCs November 8, 2023, Request for Additional Information - Related to Independent Spent Fuel Storage Installation CNL-23-068, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation CNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) ML23346A1222023-12-12012 December 2023 Annual Non-Radiological Environmental Operating Report - 2023 IR 05000327/20234202023-11-28028 November 2023 Security Baseline Inspection Report 05000327/2023420 and 05000328/2023420 CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23324A4362023-11-0909 November 2023 Exam Corporate Notification Letter Aka 210-day Letter ML23307A0822023-11-0808 November 2023 Request for Additional Information August 4, 2022, Exemption Request for Deviating from the Conditions of Certificate of Compliance No. 1032, Amendment No. 3, Related to Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation IR 05000327/20230032023-11-0303 November 2023 Integrated Inspection Report 05000327/2023003 and 05000328/2023003 ML23306A1592023-11-0202 November 2023 Enforcement Action EA-22-129 Inspection Readiness Notification ML23292A0792023-10-19019 October 2023 Tennessee Valley Authority - Emergency Plan Implementing Procedure Revision, Includes EPIP-5, Revision 58, General Emergency IR 05000327/20230112023-10-16016 October 2023 Triennial Fire Protection Inspection Report 05000327/2023011 and 05000328/2023011 ML23285A0882023-10-12012 October 2023 Submittal of Sequoyah Nuclear Plant, Units 1 and 2, Submittal of Updated Final Safety Analysis Report Amendment 31 ML23284A4252023-10-11011 October 2023 10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; Commitment Summary Report; and Update to the Fire Protection Report ML23283A2792023-10-10010 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Requirements Manual ML23279A0612023-10-0505 October 2023 Paragon Energy Solutions LLC, Part 21 Final Report Re Potential Defect with Eaton Jd and Hjd Series Molded Case Circuit Breakers (Mccbs) ML23277A0462023-10-0404 October 2023 Revisions to the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification Bases ML23275A0272023-09-29029 September 2023 Submittal of Discharge Monitoring Report (DMR) Quality Assurance Study 43 Final Report 2023 ML23271A1662023-09-28028 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision IR 05000327/20234032023-09-14014 September 2023 Cyber Security Inspection Report 05000327/2023403 and 05000328/2023403 (Cover Letter) ML23257A0062023-09-14014 September 2023 Enforcement Action EA-22-129 Inspection Postponement Request ML23254A2192023-09-11011 September 2023 Emergency Plan Implementing Procedure Revisions ML23254A0652023-09-0707 September 2023 Registration of Spent Fuel Storage Cask Pursuant to 10 CFR 72.212(b)(2) CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000327/20230052023-08-29029 August 2023 Updated Inspection Plan for Sequoyah Nuclear Plant, Units 1 and 2 - Report 05000327/2023005 and 05000328/2023005 ML23233A0122023-08-17017 August 2023 Unit 1 Cycle 25 Refueling Outage - 90-Day Inservice Inspection Summary Report - Supplement ML23233A0142023-08-15015 August 2023 Discharge Monitoring Report (Dmr), July 2023 ML23215A1212023-08-0303 August 2023 301 Exam Administrative Items (2B) Normal Release ML23215A1572023-08-0303 August 2023 Enforcement Action EA-22-129 Inspection Readiness Notification CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information 2024-01-04
[Table view] Category:Safety Evaluation
MONTHYEARML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues ML23072A0652023-04-0505 April 2023 Units 1 and 2 Issuance of Amendment Nos. 364 and 358; 160 and 68 Regarding a Revision to Technical Specification 3.4.12 ML23048A3042023-03-0808 March 2023 Tennessee Valley Authority - Request for Relief from Requirements of ASME Boiler and Pressure Vessel Code Regarding Weld Examination Coverage (EPID L-2022-LLR-0045,-0046,-0047) ML22334A0732023-02-0606 February 2023 Issuance of Amendment Nos. 363 and 357 Regarding Modification of the Approved Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors ML22348A0052023-01-25025 January 2023 Issuance of Amendment Nos. 326, 349, and 309; 363 and 35; 159 and 67 Regarding Adoption of TSTF-554, Revise Reactor Coolant Leakage Requirements ML22349A6472023-01-20020 January 2023 Issuance of Amendment Nos. 325, 348, and 308; 362 and 356; and 158 and 66 Regarding Adoption of TSTF-529, Rev. 4, Clarify Use and Application Rules ML22271A9142022-12-0707 December 2022 Issuance of Amendment Nos. 324, 347, and 307; 360 and 354; 157 and 65 Regarding a Revision to the Emergency Action Level Scheme ML22304A1862022-12-0101 December 2022 Revised Authorization of Alternative Request RV-02 for Pressure Isolation Valve Seat Leakage ML22284A0072022-11-0707 November 2022 Authorization of Alternative Request RP-12 for Testing of Centrifugal Charging Pump 1B-B ML22276A1612022-10-24024 October 2022 Issuance of Amendment Nos. 359, 353, 155, & 63 Regarding Adoption of TSTF Traveler TSTF-577, Revised Frequencies for Steam Generator Tube Inspections ML22272A5682022-10-12012 October 2022 Authorization of Alternatives to Certain Inservice Testing Requirements in the American Society of Mechanical Engineers Operating and Maintenance Code ML22259A2042022-09-29029 September 2022 Correction to Amendment Nos. 358 and 352 Regarding Technical Specifications Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B ML22263A3752022-09-29029 September 2022 Authorization of Alternative Request RV-02 for Pressure Isolation Valve Seat Leakage ML22210A1182022-08-24024 August 2022 Issuance of Amendment Nos. 358 and 352 Regarding Technical Specifications Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B ML22165A1052022-07-12012 July 2022 Issuance of Amendment Nos. 357 and 351 Regarding Revision to Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation ML22084A0012022-04-0505 April 2022 Clinch River Nuclear Site; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Units 1 and 2, Review of Quality Assurance Plan Changes ML21298A0312021-10-27027 October 2021 Issuance of Exigent Amendment No. 350 Regarding One-Time Change to Technical Specification 3.4.12, Low Temperature Overpressure Protection System, ML21245A2672021-10-27027 October 2021 NON-PROPRIETARY - Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 356 and 349 Regarding the Transition to Westinghouse Robust Fuel Assembly-2 (RFA-2) Fuel ML21266A3832021-10-0101 October 2021 Authorization of Alternative Request RP-10 for the 1B-B Motor Driven Auxiliary Feedwater Pump ML21131A1632021-05-28028 May 2021 Authorization and Safety Evaluation for Alternative Request No. 21-ISI-1 ML21084A1902021-05-0404 May 2021 Issuance of Amendment Nos. 355 and 348 Regarding Revision to Technical Specification Table 3.3.3-1, Post Accident Monitoring Instrumentation ML21021A3492021-03-0303 March 2021 Issuance of Amendment Nos. 354 and 347 Regarding Revision to Technical Specification 4.2.2, Control Rod Assemblies ML20337A0372021-02-0101 February 2021 Issuance of Amendment No. 353 Regarding Steam Generator Tube Inspection Frequency ML20350B4932021-01-25025 January 2021 Issuance of Amendment Nos. 352, 346, 141, and 47 Regarding the Adoption of Technical Specification Task Force Traveler, TSTF-569, Revision 2, Revise Response Time Testing Definition ML20268A0822021-01-12012 January 2021 Issuance of Amendment Nos. 314, 337, and 297; 351 and 345; 140 and 46 Regarding Changes to the Technical Specifications ML20282A3452020-11-19019 November 2020 Issuance of Amendment Nos. 313, 336, 296, 350, 344, 138, and 44 Revise Emergency Plan On-Shift Emergency Medical Technician & Onsite Ambulance Requirements ML20262H0262020-11-12012 November 2020 Issuance of Amendment Nos. 349 and 343 Regarding Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure ML20108F0492020-04-23023 April 2020 Issuance of Amendment No. 342 Exigent Amendment to Operate One Cycle with One Control Rod Removed ML19319C8312019-11-21021 November 2019 Issuance of Exigent Amendment No. 348 to Operate One Cycle with One Control Rod Removed ML19281B5542019-11-18018 November 2019 Issuance of Amendment Nos. 347 and 341 Request to Adopt TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program ML19234A0132019-09-23023 September 2019 Alternative Request 18-ISI-1 Regarding Examination of Dissimilar Metal Welds in Reactor Vessel Head ML19179A1352019-09-18018 September 2019 Issuance of Amendment Nos. 346 and 340, Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors ML18277A1102019-08-27027 August 2019 Units, 1 & 2 Issuance of Amendment Nos. 309, 332, 292, 345, 339, 128, and 31 Regarding Unbalanced Voltage Protection ML19227A1102019-08-26026 August 2019 Alternative Request for the Turbine Driven Auxiliary Feedwater Pumps 10-Year Interval Inservice Testing Program ML19196A2212019-07-18018 July 2019 Issuance of Exigent Amendment Technical Specification Change - Reactor Vessel Level Instrumentation Inoperable ML19058A0292019-05-0707 May 2019 Issuance of Amendments Request to Modify Essential Raw Cooling Water Control Center Breakers and to Revise the Updated Final Safety Analysis Report (SQN-TS-17-04) ML19010A2742019-03-18018 March 2019 Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report- TVA Overall Basin Probable Maximum Precipitation and Local Intense Precipitation, Calculation, CDQ0000002016000041 Tennessee Valley Authority ML18197A3072018-08-30030 August 2018 Issuance of Amendments Regarding Request to Change Technical Specification 3.3.1 and Surveillance Requirement 3.2.4 ML18159A4612018-08-0606 August 2018 Issuance of Amendments Regarding Request to Change Emergency Plan ML18138A4522018-05-29029 May 2018 Bf, Units 1, 2, and 3; Sequoyah, Units 1 and 2; Watts Bar, Units 1 and 2 - Correction to an Omitted Reference for License Amendment Regarding Request to Upgrade (CAC Nos. MF9054, MF9055, MF9056, MF9057, MF9058, MF9059, and MF9060, EPID L-20 ML17289A0322017-12-22022 December 2017 Issuance of Amendments Regarding Request to Upgrade Emergency Action Level Scheme (CAC Nos. MF9054-60; EPID L2017-LLA-0160) ML17034A3602017-03-27027 March 2017 Issuance of Amendment Nos. 298, 322, 282, and 338 and 331 - Revise Technical Specification 5.3, Unit Staff Qualifications to Replace References to Rg 1.8, Rev. 2 with TVA Nuclear Quality ... ML16340A0092017-03-0909 March 2017 Safety Evaluation Regarding Implementation of Mitigation Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML17059B7912017-03-0202 March 2017 Sequoyah Nuclear Plant, Units 1 and 2 - Relief from the Requirements of the American Society of Mechanical Engineer OM Code (CAC Nos. MF9305 and MF9306) ML16228A0962016-10-0303 October 2016 Issuance of Amendments to Change the Completion Date of Cyber Security Plan Implementation Milestone 8 ML16225A2762016-09-29029 September 2016 Issuance of Amendments to Revise Technical Specification for Essential Raw Cooling Water System Allowed Completion Time ML16225A6332016-09-0202 September 2016 Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Unit 1 and 2 - Relief Request for Use of Alternate Calibration Block Reflector Requirements 16-PDI-5 (TAC Nos. MF7754-MF7760) ML16123A1312016-05-12012 May 2016 Relief Requests RP-01, RP-02, RP-06, RP-08, and RV-01 Related to the Inservice Testing Program, Fourth 10-Year Interval (CAC Nos. MF7099 and MF7100) ML15329A1862015-12-0404 December 2015 Request for Relief PR-07 for Alternative Inservice Pump Testing at Reference Values ML15320A2182015-11-30030 November 2015 Issuance of Amendments to Revise Technical Specification for Containment Leakage Rate Program 2024-01-29
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UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 December 22, 2011 Mr. Joseph W. Shea Corporate Manager, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SEOUOYAH NUCLEAR PLANT, UNIT 2 -SAFETY EVALUATION OF RELIEF REOUEST NO. 2-APPJ-1, PROPOSED ALTERNATIVE TO THE LEAKAGE TESTING REOUIREMENTS IN PARAGRAPH IWE-S221 OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI, 2001 EDITION WITH 2003 ADDENDA (TAC NO. MES24S) Dear Mr. Shea: By letter dated December 22,2010 (Agencywide Documents Access and Management System Accession No. ML 103630044) Tennessee Valley Authority (TVA), the licensee, submitted Relief Request 2-APPJ-1 for the Sequoyah Nuclear Plant (SON), Unit 2. Pursuant to Title 10 of the Code ofFederal Regulations (10 CFR) SO.SSa(a)(3)(i), TVA requested to utilize an alternative to the testing requirements in paragraph IWE-S221,Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2001 edition with the 2003 addenda. TVA's request for relief proposes alternative test methods (pneumatic leakage test) for containment vessel repair activities following the SON, Unit 2 Cycle 18 steam generator replacement refueling outage in fall 2012. The proposed alternative to the requirements of paragraph IWE-S221 will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR SO.SSa(a)(3(i>>, the Nuclear Regulatory Commission authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2.
J. W. Shea All other requirements of the ASME Code, Sections III and XI, for which relief has not been specifically requested and approved remain applicable. If you have any questions, please feel free to contact the SON Project Manager, Siva P. Lingam, at (301) 415-1564. Sincerely,/4;:;70/.. .. /J ) -Ii*****--** (crY Douglas A. Broaddus, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328 Enclosure: Safety Evaluation cc wfencl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 2-APPJ-1 ALTERNATIVE TEST METHODS FOR CONTAINMENT REPAIR ACTIVITIES TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 DOCKET NO. 50-328 1.0 INTRODUCTION By the letter dated December 22, 2010 (Agencywide Documents Access and Management System Accession No. ML103630044), Tennessee Valley Authority (TVA, the licensee) proposed an alternative, which would allow the licensee to perform a local leak rate test of the containment in lieu of the "Leakage Test" requirements in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Code (ASME Code). The leakage tests are required as a result of planned repairs to the Sequoyah Nuclear Plant (SQN), Unit 2 steel containment vessel (SCV) in accordance with Title 10 of the Code ofFederal Regulations (10 CFR), 50.55a(a)(3)(i). The licensee will be replacing the SQN, Unit 2 steam generators during the cycle 18 refueling outage, commencing in the fall 2012. This safety evaluation addresses the ability of the proposed alternative to ensure the continued ability of the SCV to provide an acceptable level of quality and safety after the steam generator replacement activity. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested whether they could use an alternative to the requirements in Section XI of the ASME Code for post repair leakage inspection of the SQN, Unit 2 SCV from the scheduled steam generator replacement project scheduled to start in October 2012. The regulations at 10 CFR 50.55a, "Codes and Standards," incorporate, by reference, the 2001 edition through the 2003 addenda of Section XI of the ASME Code. Paragraph IWE-5221 of Subsection IWE of the ASME Code,Section XI, requires a leakage rate test following any repair and replacement activity. Paragraph IWE-5221 specifies that the leakage rate test be conducted in accordance with the provisions of 10 CFR Part 50, Appendix J, paragraph IV.A, "Containment modification," which states in part, "Any major modification, replacement of a component, which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable for the area affected by the modification,"
3.0 TECHNICAL EVALUATION
3.1 Specific Alternative Request The licensee proposes to perform an "as-left" local leak test on the new pressure boundary welds of the SCV repair welds in lieu of the Type A, Type B, or Type C leakage rate tests as specified in Paragraph IWE-5221 of the ASME Code for this type of repair activity. 3.2 Proposed Alternative Duration The duration of the proposed alternative is through the completion and approval of all testing associated with the restoration of the containment opening created to support the SON, Unit 2 steam generator replacement. 3.3 Basis for Relief The licensee committed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of Section III, ofthe ASME Code, 1968 Edition through the winter 1968 addenda. Before performing the repair weld, the surfaces to be welded will be cleaned. A magnetic particle test of the back gouge of the root pass will be performed, and 100 percent radiography will be performed on the pressure boundary of SCV final repair welds. In addition, ASME Code,Section XI, as modified by 10 CFR 50.55a(b)(2)(ix)(G), requires both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12.0 pounds per square inch gauge (psig) for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated. Based on the above procedure, the licensee stated that any local leakage will be directly identified by the bubble test and it will confirm the findings of the prior nondestructive examination tests. The combination of the 100 percent radiography, demonstrating that the repair welds meet the construction code radiography acceptance criteria; and the local leak rate test of the repair welds, by performing the bubble test while the SCV and repair welds are at accident pressure, will provide a more effective examination than the Type A test, as required by the ASME Code, paragraph IWE-S221. Therefore, an alternative to the requirement of paragraph IWE-S221 is requested pursuant to 10 CFR SO.55a(a)(3)(i), in that the proposed alternative provides an acceptable level of quality and safety. 3.4 Nuclear Regulatory Commission (NRC) Staff Evaluation To facilitate the SON, Unit 2 steam generator replacement, the SCV will be breached. An opening will be cut in the SCV in order to remove and replace the steam generators. After the steam generator replacement, the SCV sections removed will be reattached through welding. Paragraph IWE-5221 of the ASME Code requires that leakage rate testing be conducted to ensure the integrity of the repairs before returning the SCV to operable status. In lieu of the Type A, Type B, or Type C leakage rate testing, the licensee proposed to perform a series of examinations and a leak test subjecting the SCV to accident pressure, to verify the leak tightness and integrity of the liner welds and the SCV.
-The licensee's proposed relief request includes the detailed examination and test sequence that is summarized herein. The licensee proposed to perform the activities described below as a part of the SCV restoration effort. The sections of the SCV that were removed will be rewelded in place in accordance with the requirements of the ASME Code,Section III, 1968 edition through the winter 1968 addenda (TVA's code of record requirements). Before performing the repair weld, the surfaces to be welded will be cleaned and examined by magnetic particle testing of the weld preparation area, and 100 percent radiography of the final repair weld will be performed. In addition, both a general visual, VT-3 examination and a detailed visual, VT-1 examination of the SCV pressure boundary welds will be conducted. To perform a weld leak test, the containment will be pressurized to a test pressure Pa of at least 12 psig for a minimum of 10 minutes. The test pressure will be held between 12.2 psig and 12.5 psig. A zero leakage criterion will be used for weld acceptance, which is determined by the absence of any bubbles formation. Any leakage identified will be corrected, and the test will be repeated The magnetic particle or liquid penetrate testing, the 100 percent radiography of the repair weld, and the subsequent bubble test will provide adequate assurance that the repair welds do not leak or have any structural defects. The zero leakage acceptance criterion for the bubble test will ensure that the SCV leakage rate is not altered by the steam generator replacement activity, and the pressurization of the SCV to the accident pressure will confirm the integrity of the SCV after the repair. Therefore, the NRC staff concludes that the proposed alternative will provide adequate assurance of structural integrity. 4.0 CONCLUSION On the basis of the information provided in the relief request, and an independent NRC staff evaluation, the NRC staff concludes that the proposed alternative of a local leak rate test in lieu of the requirements in paragraph IWE-5221 of the ASME Code will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3(i)), the NRC authorizes the use of the proposed one-time alternative request during the licensee's steam generator replacement program, which is scheduled during the Cycle 18 refueling outage in fall 2012 for SON, Unit 2. All other requirements of the ASME Code,Section III and Section XI, for which relief has not been specifically requested and approved, remain applicable. Principal Contributor: D. Hoang Date: December 22,2011 J. W. Shea -2 All other requirements of the ASME Code, Sections III and XI, for which relief has not been specifically requested and approved remain applicable. If you have any questions, please feel free to contact the SON Project Manager, Siva P. Lingam, at (301) 415-1564. Sincerely, IRA by TOrfforl Douglas A. Broaddus, Chief Plant licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Distribution: PUBLIC LPL2-2 r/f RidsOgcMailCenter RidsNrrDorlDpr RidsAcrsAcnw MailCTR RidsNrrDorlLPL2-2 RidsNrrLACSola RidsNrrPMSequoyah RidsRgn2MailCenter RidsNrrDeEmcb RidsNrrDssScvb D. Hoang, NRR ADAMS Accession No.: ML11313A165 *by memo dated OFFICE LPL2-2/PM LPL2-2/LA EMCB/BC SCVB/BC NAME SLingam RSoia MKhanna* RDennig DBroaddus DATE 11/29/11 11/29/11 10/26/11* 12/7/11 12/22/11 OFFICIAL RECORD COPY