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MONTHYEARML14013A3022014-01-10010 January 2014 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&Pv) Section XI Code of Record for Second 10-Year Lnservice Interval 2001 Edition Project stage: Request ML14017A0972014-01-17017 January 2014 Accepted for Review of Watts Bar Nuclear Plant, Unit 1, Request for an Alternative System Leakage Test (Rfa) ISPT-02, TAC MF3354 Project stage: Approval ML14042A3212014-02-11011 February 2014 E-mail - Request for Additional Information Related to Request for Alternative Number ISPT-02 Regarding System Leakage Test of Reactor Pressure Vessel Head Flange Seal Leak-off Piping Project stage: RAI ML14055A2712014-02-21021 February 2014 Response to Request for Additional Information Regarding American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel(B&Pv) Section Xl Code of Record for Second 10-Year Inservice Interval 2001 Edition Project stage: Response to RAI ML14073A6002014-03-13013 March 2014 Response to Request for Clarifying Lnformation Regarding American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&Pv) Section Xl Code of Record for Second 10-Year Lnservice Lnterval 2001 Edition Project stage: Request ML14079A4772014-03-27027 March 2014 Request for Alternate ISPT-02 Regarding System Leakage Test of Reactor Pressure Vessel Head Flange Seal Leak-Off Piping Project stage: Other 2014-02-11
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Category:Letter
MONTHYEARCNL-24-060, Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description2024-09-24024 September 2024 Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description CNL-24-047, Decommitment of Flood Mode Mitigation Improvement Systems2024-09-24024 September 2024 Decommitment of Flood Mode Mitigation Improvement Systems ML24262A0602024-09-23023 September 2024 Summary of August 19, 2024, Meeting with Tennessee Valley Authority Regarding a Proposed Supplement to the Tennessee Valley Authority Nuclear Quality Assurance Plan CNL-24-065, Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-09-18018 September 2024 Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions 05000390/LER-2024-002, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation2024-09-0505 September 2024 Automatic Reactor Trip Due to Main Generator Protection Relay Actuation IR 05000390/20240052024-08-28028 August 2024 Updated Inspection Plan and Assessment Follow-Up Letter for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390-2024005 and 05000391-2024005 ML24218A1442024-08-27027 August 2024 Issuance of Amendment Nos. 169 and 75 Regarding Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate IR 05000390/20244022024-08-20020 August 2024 – Security Baseline Inspection Report 05000390-2024402 and 05000391/2024402 - Public CNL-24-061, Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08),2024-08-19019 August 2024 Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08), ML24219A0262024-08-12012 August 2024 Request for Withholding Information from Public Disclosure IR 05000390/20240022024-08-0707 August 2024 Integrated Inspection Report 05000390/2024002 and 05000391/2024002 Rev ML24204A2652024-07-25025 July 2024 Regulatory Audit Summary Related to Request to Revise Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24170A8002024-07-15015 July 2024 Issuance of Amendment Nos. 168 and 74 Regarding Revision to Technical Specification Table 1.1-1 for Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts IR 05000390/20244402024-07-12012 July 2024 95001 Supplemental Inspection Supplemental Report 05000390-2024440 and 05000391-2024440 and Follow-Up Assessment Letter ML24131A0012024-07-0202 July 2024 Issuance of Amendment Nos. 167 and 73 Regarding Adoption of Technical Specification Task Force Traveler TSTF-427-A, Revision 2 CNL-24-052, Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-06-27027 June 2024 Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-24-018, License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS2024-06-25025 June 2024 License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc ML24100A7642024-05-16016 May 2024 Issuance of Amendment No. 166 Regarding Revision to Technical Specification 3.8.2, AC Sources-Shutdown, to Remove Reference to C-S Diesel Generator (CNL-23-062) IR 05000390/20240012024-05-14014 May 2024 Integrated Inspection Report 05000390/2024001 and 05000391/2024001 CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000391/20240072024-04-30030 April 2024 Assessment Follow-up Letter for Watts Bar Nuclear Plant, Unit 2 – Report 05000391/2024007 ML24120A1182024-04-29029 April 2024 – Notification of NRC Supplemental Inspection (95001) and Request for Information CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 ML24087A1912024-04-18018 April 2024 Exemption from Select Requirements of 10 CFR Part 73, Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting CNL-24-010, License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19)2024-04-17017 April 2024 License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19) CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24072A0052024-04-15015 April 2024 Issuance of Amendment Nos. 165 and 72 Regarding Increase in the Maximum Number of Tritium Producing Burnable Absorber Rods and Supporting Changes, and Revision to the Updated Final Safety Analysis Report CNL-24-004, Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13)2024-04-0404 April 2024 Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13) IR 05000390/20244012024-04-0202 April 2024 – Security Baseline Inspection Report 05000390/2024401 and 05000391/2024401 - (Public) CNL-24-020, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements2024-04-0101 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements CNL-24-007, Annual Insurance Status Report2024-03-27027 March 2024 Annual Insurance Status Report CNL-24-008, Guarantee of Payment of Deferred Premiums - 2023 Annual Report2024-03-27027 March 2024 Guarantee of Payment of Deferred Premiums - 2023 Annual Report CNL-24-025, Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule2024-03-25025 March 2024 Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule ML24081A0262024-03-21021 March 2024 Emergency Plan Implementing Procedure Revisions ML24079A0312024-03-19019 March 2024 Wb 2024-301, Corporate Notification Letter (210-day Ltr) CNL-24-031, Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-03-18018 March 2024 Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance CNL-24-028, Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2024-03-14014 March 2024 Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation CNL-24-029, Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 372024-03-14014 March 2024 Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 37 IR 05000390/20240102024-03-0808 March 2024 Age-Related Degradation Inspection Report 05000390/2024010 and 05000391/2024010 CNL-24-012, Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-02-28028 February 2024 Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance IR 05000390/20230062024-02-28028 February 2024 Annual Assessment Letter for Watts Bar Nuclear Plant Units 1 and 2 - Report 05000390/2023006 and 05000391/2023006 CNL-24-023, Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 252024-02-20020 February 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 25 ML24009A1712024-02-16016 February 2024 Environmental Assessment and Finding of No Significant Impact Related to an Increase in the Maximum Number of Tritium Producing Burnable Absorber Rods (EPID L-2023-LLA-0039) - Letter ML24019A0722024-02-14014 February 2024 Request for Withholding Information from Public Disclosure IR 05000390/20230042024-02-13013 February 2024 Integrated Inspection Report 05000390/2023004 and 05000391/2023004 and Apparent Violation 2024-09-05
[Table view] Category:Safety Evaluation
MONTHYEARML24218A1442024-08-27027 August 2024 Issuance of Amendment Nos. 169 and 75 Regarding Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate ML24170A8002024-07-15015 July 2024 Issuance of Amendment Nos. 168 and 74 Regarding Revision to Technical Specification Table 1.1-1 for Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts ML24100A7642024-05-16016 May 2024 Issuance of Amendment No. 166 Regarding Revision to Technical Specification 3.8.2, AC Sources-Shutdown, to Remove Reference to C-S Diesel Generator (CNL-23-062) ML24072A0052024-04-15015 April 2024 Issuance of Amendment Nos. 165 and 72 Regarding Increase in the Maximum Number of Tritium Producing Burnable Absorber Rods and Supporting Changes, and Revision to the Updated Final Safety Analysis Report 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 ML23293A0572023-12-0606 December 2023 Issuance of Amendment Nos. 163 and 70 Regarding Adoption of TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML23233A0042023-08-28028 August 2023 Proposed Alternative to the Requirements of the ASME Boiler and Pressure Vessel Code for Upper Head Injection Dissimilar Metal Butt Welds ML23125A2202023-05-0505 May 2023 Issuance of Amendment No. 161 Regarding a Change to Footnotes for Technical Specification Table 1.1-1 Modes (Emergency Circumstances) 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) 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 ML22293A4082022-11-14014 November 2022 Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule ML22257A0512022-11-0404 November 2022 Issuance of Amendment Nos. 156 and 64 Regarding Adoption of TSTF-205-A, Revision 3, and TSTF-563-A 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 ML22187A1812022-09-20020 September 2022 Issuance of Amendment Nos. 153 and 62 Regarding Extension of Completion Time for Technical Specification 3.7.8 for Inoperable Essential Raw Cooling Water Train ML22187A0192022-09-20020 September 2022 Issuance of Amendment No. 154 Regarding Revision to Technical Specification 3.3.2 to Revise Allowable Value for Trip of Turbine-Driven Main Feedwater Pumps ML22014A2062022-05-0404 May 2022 Issuance of Amendment Nos. 152 and 61 Regarding Revision to Technical Specifications to Delete a Redundant Unit of Measure for Certain Radiation Monitors 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 ML22070A0022022-03-28028 March 2022 Review of the Fall 2021 Mid Cycle Generic Letter 95-05 Voltage-Based Alternate Repair Criteria Report ML21306A2872022-01-25025 January 2022 Issuance of Amendment No. 60 Regarding Revision of Technical Specification Requirements Specific to the Model D3 Steam Generators That Will No Longer Apply Following Steam Generator Replacement ML21334A2952022-01-18018 January 2022 Issuance of Amendment No. 151 Regarding Revision to TS 3.7.12 for One-Time Exception to Permit Continuous Opening of Auxiliary Building Secondary Containment Enclosure During Unit 2 Steam Generator Replacement ML21334A3892022-01-12012 January 2022 Issuance of Amendment No. 59 Regarding Revision to Steam Generator Tube Rupture Dose Analysis ML21271A1372021-12-16016 December 2021 Issuance of Amendment Nos. 150 and 58 Regarding Modification of Technical Specification Surveillance Requirement 3.6.15.4, Shield Building ML21260A2102021-11-22022 November 2021 Issuance of Amendment No. 57 to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level ML21189A3072021-11-0303 November 2021 Issuance of Amendment Nos. 149 and 56 Regarding Modification of Technical Specification 5.7.2.19, Containment Leakage Rate Testing Program ML21158A2842021-09-17017 September 2021 Issuance of Amendment Nos. 148 and 55 to Revise Technical Specifications for Function 6.E of Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation ML21153A0492021-07-26026 July 2021 Issuance of Amendment No. 147 Regarding Change to Steam Generator Tube Inspection Frequency and Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-510 ML21161A2392021-06-24024 June 2021 Issuance of Amendment No. 54 Regarding Use of Temperature Adjustment to Voltage Growth Rate for the Generic Letter 95-05 Steam Generator Tube Repair Criteria ML21148A1002021-06-17017 June 2021 Issuance of Amendment No. 53 Regarding Neutron Fluence Calculation Methodology ML21099A2462021-05-14014 May 2021 Issuance of Amendment Nos. 146 and 52 to Adopt TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec ML21130A6012021-05-13013 May 2021 Correction of Proposed Alternative IST-RR-8 to the Requirements of the ASME OM Code for the Residual Heat Removal Pump 1B-B ML21078A4842021-05-0505 May 2021 Issuance of Amendment Nos. 145 and 51 for One-Time Change to Technical Specification 3.7.11 to Extend the Completion Time for Main Control Room Chiller Modifications ML21110A0372021-04-29029 April 2021 Proposed Alternative IST-RR-8 to the Requirements of the ASME OM Code for the Residual Heat Removal Pump 1B-B ML21064A4082021-03-10010 March 2021 Correction of Safety Evaluation for License Amendment Nos. 143 and 50 (EPID L-2020-LLA-0005) (Non-Proprietary) ML21015A0342021-03-0909 March 2021 Issuance of Amendment No. 144 Regarding Post Accident Monitoring Instrumentation ML21034A1692021-02-26026 February 2021 Issuance of Amendment Nos. 143 and 50 Regarding Implementation of Full Spectrumtm Loss-of-Coolant Accident Analysis (LOCA) and New LOCA-Specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology ML20232C6222021-02-11011 February 2021 Issuance of Amendment Nos. 142 and 49 Regarding Revision to Technical Specifications to Implement WCAP-17661-P-A, Revision 1, Improved RAOC and CAOC Fq Surveillance Technical Specifications (EPID L-2020-LLA-0037 ML21027A1672021-02-0909 February 2021 Issuance of Amendment No. 48 Regarding Use of Alternate Probability of Detection Values for Beginning of Cycle in Support of Operational Assessment 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 ML20245E4132020-12-0808 December 2020 Issuance of Amendment Nos. 139 and 45 Regarding Revisions to Technical Specification 3.6.15, Shield Building 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 ML20239A7912020-10-28028 October 2020 Issuance of Amendment Nos. 137 and 43 Regarding Revision to Technical Specifications to Adopt Technical Specification Task Force Traveler 541, Revision 2 ML20226A4442020-10-21021 October 2020 Issuance of Amendment No. 42 Regarding Measurement Uncertainty Recapture Power Uprate ML20167A1482020-08-19019 August 2020 Issuance of Amendment Nos. 136 and 41 Regarding the Automatic Transfer from a Unit Service Station Transformer to a Common Station Service Transformer ML20156A0182020-08-10010 August 2020 Issuance of Amendment No. 40 Regarding Technical Specifications for Steam Generator Tube Repair Sleeve ML20076A1942020-04-30030 April 2020 Issuance of Amendment Nos. 134 and 38 Regarding Adopting the Title 10 CFR Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants 2024-08-27
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 27, 2014 Christopher Church, Vice President Tennessee Valley Authority Watts Bar Nuclear Plant P.O. Box 2000 Spring City, TN 37381 SUBJECT: WATTS BAR NUCLEAR PLANT, UNIT NO. 1 *REQUEST FOR ALTERNATIVE ISPT*02 REGARDING SYSTEM LEAKAGE TEST OF REACTOR PRESSURE VESSEL HEAD FLANGE SEAL LEAK*OFF PIPING (TAC NO. MF3354) Dear Mr. Church: By letter dated January 10, 2014, as supplemented by letters dated February 21, 2014 and March 13, 2014, Tennessee Valley Authority (TVA, the licensee) submitted to the Nuclear Regulatory Commission (NRC) a request for alternatives (RFA) ISPT[Inspection System Pressure Testing]*02 to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI [or ASME Code for Operation and Maintenance of Nuclear Power Plants {OM Code)] requirements at Watts Bar Nuclear Plant, Unit No.1. Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.55a(a)(3)(ii), the licensee proposed an alternative system leakage test of the reactor pressure vessel head flange seal leak*off piping on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that TVA has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a{a)(3)(ii). Therefore, the NRC staff authorizes the use of RFA ISPT*02 at Watts Bar, Unit 1, for the remainder of the second 1 0-year lnservice Inspection interval that commenced on May 27, 2007, and will end on May 26, 2016. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.
C. Church -2-If you have any questions, please contact the Project Manager, Andrew Hen at 301-415-8480 or via e-mail at Andrew.Hon@nrc.gov.
- Docket No. 50-390 Enclosure: Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, . s1e . Q ichocho, Chief lant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE ISPT-02 REGARDING SYSTEM LEAKAGE TEST OF REACTOR PRESSURE VESSEL HEAD FLANGE SEAL LEAK-OFF PIPING 1.0 INTRODUCTION TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NUMBER 50-390 By letter dated January 10, 2014 (Agencywide Documents Access and Management System Accession No. ML 14013A302), as supplemented by letters dated February 21, 2014 (Accession No. ML 14055A271 ), and March 13, 2014 (Accession No. ML 14073A600), Tennessee Valley Authority (the licensee) submitted for the U.S. Nuclear Regulatory Commission (NRC) approval request for alternative (RFA) ISPT[Inspection System Pressure Testing]-02. The licensee proposed an alternative to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. The request relates to the inservice inspection (lSI) requirement of Sub-Article IWC-5221 when the licensee performs a system leakage test of the reactor pressure vessel (RPV) head flange seal leak-off piping. Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.55a(a)(3)(ii), the licensee proposed an alternative system leakage test of the RPV head flange seal leak-off piping, given the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(g)(4), the ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The inservice examination of components and system pressure tests conducted during the first 1 0-year interval and subsequent intervals must comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(b), 12 months prior to the start of the 120-month interval, subject to the conditions listed therein. Enclosure
-2-Pursuant to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee. 3.0 TECHNICAL EVALUATION 3.1 The Licensee's Request for Alternative The component affected by this request is the RPV head flange seal leak-off detection piping (1-PIPE-68-B) and (1-47W813-1) extending from the two RPV taps and terminating at the inline isolation valve 1-FCV-68-22 of the common header to the reactor coolant drain tank. The subject piping is classified as a pressure retaining component under ASME Code Class 2, IWC-2500, Table IWC-2500-1, Examination Category C-H, Item Number C7.1 0. The RPV head flange seal leak detection line consists of 21 feet of 3/4 inch and 3/8 inch piping, and less than 12 inches of 1 inch piping. The design conditions for the pipe are 2500 pound per square inch absolute at 650 degrees Fahrenheit (°F). In a letter dated February 21, 2014, the licensee stated that the piping was manufactured from stainless steel SA376, TP304 material, and is Schedule 160. The ASME Code of Record for the second 1 0-year lSI interval at Watts Bar Unit 1, is the 2001 Edition through 2003 Addenda of the ASME Code. The ASME Code,Section XI, IWC-2500, Table IWC-2500-1, Examination Category C-H, establishes requirements to conduct the system leakage testing according to IWC-5220 and the VT-2 visual examination according to IWA-5240 during each inspection period. As required by IWC-5221, the system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability. The licensee proposed an alternative to IWC-5221. In lieu of normal operating pressure, the licensee proposed to use the static pressure head of 11 pounds per square inch gauge (psig), developed at the taps from the elevation of 25.5 feet of water above the reactor vessel closure flange when the cavity is flooded, to perform the system leakage testing of the leak-off piping. The licensee stated that the VT -2 visual examination of the accessible portions (from the biological shield wall to 1-FCV-68-22) of the leak-off piping will be performed at ambient conditions when the RPV head is off and the reactor cavity is flooded above the RPV flange for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The VT -2 visual examination of the inaccessible portions (from the biological shield wall to the RPV flange taps) of the leak-off piping will be performed by observing the area near the leak-off line for evidence of leakage in conjunction with the visual inspection of the RPV hot leg nozzle safe end welds. The licensee in a letter dated March 13, 2014, stated that in order to reduce the dose associated with the hot leg inspection and the
-3 -VT -2 visual examination of the leak-off pipe, the same personnel will be utilized to perform both examinations simultaneously while making a single entry inside the biological shield wall. Both the hot leg inspection and the VT -2 visual examination of the inaccessible area of the leak-off pipe are scheduled during the core off-load period with the reactor cavity flooded to the normal refueling level that provided the same static pressure head. The licensee stated that the VT -2 visual examination of the inaccessible portion of the RPV flange leak-off line will be able to detect evidence of boric acid residue, staining, or any other indications of past leakage. The licensee stated that the RPV head flange leak detection line is separated from the reactor pressure boundary by one 0-ring (inner 0-ring) located on the RPV flange. A second 0-ring (outer 0-ring) is located on the opposite side of the tap in the RPV flange. The line that connects to the inner 0-ring is required during plant operation to detect leakage from and failure of the inner flange seal 0-ring. The licensee continually monitors for leakage past the RPV head flange seal leak detection line 0-rings using the plant instrumentation (temperatures in excess of 140 oF). The licensee stated that the configuration of the RPV head flange leak detection lines precludes manual system leakage testing while the RPV head is removed. The configuration of the RPV flange taps combined with the small size of the taps and the high test pressure requirement prevents the taps from being temporarily plugged. Plugging or installing a pressure connection would require machining threads in each flange opening with attendant concern over chips from machining that may become a foreign material threat for the fuel integrity or the lines. Additionally, machining would require extensive time, and the radiation level at the RPV flange is estimated at 20 to 40 milliroentgen equivalent man per minute (mrem/min). After machining the threads, installation and removal of the plugs and pressure connections would require additional time for personnel to be at the radiation area with estimated dose rate of 20 to 40 mrem/min. The licensee stated that while the RPV head is installed, an adequate pressure test cannot be performed because the inner 0-ring is designed to withstand pressure in one direction only. Pressurization in the opposite direction to perform the required leak test would likely damage the inner 0-ring. Purposely failing the inner 0-ring to perform the ASME Code leakage test would require additional time to de-tension the RPV head, install the new 0-rings, and reset and re-tension the RPV head. The licensee stated in a letter dated February 21, 2014, that the radiation dose for these activities is estimated to be 1 to 1.5 roentgen equivalent man per hour. In addition, attempt to pressure test the lines in the beginning of a refueling outage when the RPV head is on would most likely result in failure of the 0-ring, which would prevent achieving the required test pressure. The licensee submitted RFA ISPT-02 for the remaining of the second 1 0-year lSI interval that commenced on May 27, 2007, and will end on May 26, 2016. The licensee stated that the second 1 0-year lSI interval was shortened to 9 years due to the first 1 0-year lSI interval being extended by 1 year to 11 years. The NRC staff notes that the reduction or extension of lSI interval within one calendar year is permitted under the ASME Code,Section XI, IWA-2430.
-4-3.2 NRC Staff Evaluation The NRC staff has evaluated RFA ISPT-02 pursuant to 10 CFR 50.55a(a)(3)(ii). The NRC staff focuse its review on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The NRC staff determined that compliance with the ASME Code,Section XI, IWC-5221, pressure requirement to perform system leakage testing of the RPV head flange seal leak-off piping would result in hardship. The basis for the hardship is as follows. The licensee would have to modify the existing flange taps during the refueling outage when the reactor head is removed. The modification of flange tap involves machining threads to install a plug or pressure connections. These activities could introduce foreign materials into the reactor pool as well as the lines. The time spent for modifications, preparing and pressurizing the lines to perform the ASME Code system leakage test, and removal of plugs or connections after completion of the test would expose personnel to additional radiation dose. Pressurizing the leak-off lines while the RPV head is installed would not be possible due to design and configuration of the RPV head flange taps, and the inner 0-ring. The inner 0-ring is designed to withstand pressure in one direction only. Therefore, the NRC staff determines that unnecessarily challenging the Foreign Material Exclusion program and the 'as low as is reasonably achievable' exposure goal constitute a hardship. The NRC staff finds that the licensee's proposed system leakage test will subject the RPV flange seal leak-off piping to the highest pressure that is obtainable without major design modifications to existing configurations of both the vessel flange face and the leak-off piping. Specifically, the static pressure head of 11 psig developed from the elevation of water above the vessel flange during the cavity flood-up will be used to pressurize the leak-off line piping to perform system leakage test. By performing the VT -2 visual examination (according to IWA-5240) of the accessible area of the leak-off lines, the licensee will be able to detect any leakage if it originated from an existing flaw in the leak-off piping and its welded connections after maintaining the static test pressure. For the inaccessible area of the leak-off lines, the licensee will use the VT-2 visual examination (according to IWA-5240) to detect evidence of any leakage such as boric acid residue and staining by observing the surrounding area of the leak-off piping. The VT -2 visual examination of the inaccessible area of the leak-off lines will be performed concurrently with the RPV hot leg nozzle safe end weld inspections because single entry inside the biological shield wall to perform both examinations will significantly reduce the accrual of radiation dose by the involved personnel. The NRC staff determined that any evidence of leakage, if it originated from an existing flaw in the leak-off piping and its welded connections will be identified by the VT-2 visual examination. The licensee stated that there has not been any documented history of degradation of the RPV flange leak-off piping within its fleet and the industry. The NRC staff review of operating experience, including Watts Bar Unit 1, did not identify any documented known degradation mechanism such as stress corrosion cracking and fatigue in the vessel flange leak-off piping and its welded connections. Furthermore, the NRC staff determined that the existing reactor coolant leakage detection systems are sufficient to provide warning to the control room operator in an unlikely event of a
-5 -through-wall leak in the RPV leak-off piping. The NRC staff finds that if the subject piping developed a through-wall flaw, the reactor coolant leakage detection systems will be able to identify the leakage during normal operation, and the licensee will take appropriate corrective actions in accordance with the plant technical specifications. Therefore, the NRC staff finds that the proposed system leakage testing using proposed test pressure is adequate to provide a reasonable assurance of structural integrity and leak tightness of the RPV flange seal leak-off piping. 4.0 CONCLUSION As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the RPV head flange seal leak-off piping. The NRC staff finds that complying with the specified ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii). Therefore, the NRC staff authorizes the use of RFA ISPT-02 at Watts Bar, Unit 1, for the remainder of the second 10-year lSI interval that commenced on May 27, 2007, and will end on May 26,2016. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third party review by the Nuclear In-service Inspector Principal Contributor: A. Rezai C. Church -2-If you have any questions, please contact the Project Manager, Andrew Hon at 301-415-8480 or via e-mail at Andrew.Hon@nrc.gov. Docket No. 50-390 Enclosure: Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION: PUBLIC LPL2-2 R/F RidsNrrDoriDpr Resource RidsNrrDorllpl2-2 Resource RidsNrrDe Resource RidsNrrDeEsgb Resource RidsNrrLABCiayton Resource RidsNrrPMWattsBar1 Resource RidsNrrPMWattsBar2 Resource RidsRgn2MaiiCenter Resource RidsAcrsAcnw _ Ma iiCTR Resources A. Rezai, NRR ADAMS A ccess1on N ML 14079A477 0. Sincerely, IRA! Jessie F. Quichocho, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation *b "I ML 14078A113 1yema1 OFFICE NRR/DORLILP2-2/PM NRR/DORLILPL2-2/LA NRR/DE/ESGB
- NRR/DORL/LPL2-2/BC NAME AHon BCiayton GKulesa JQuichocho DATE 3/21/14 3/21/14 03/19/2014 3/27/14 OFFICIAL RECORD COPY