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Category:Code Relief or Alternative
MONTHYEARML20143A2332020-06-0303 June 2020 Relief Request for Deferral of ASME Boiler and Pressure Vessel Code, Section XI, Required Summary Report Submittal (EPID L-2020-LLR-0074 (COVID-19)) ML20034E8942020-02-11011 February 2020 Relief Request Gen-ISI-ALT-2019-01 for Proposed Alternative to Use of ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 ML19361A0562020-01-0909 January 2020 Proposed Alternative HNP-ISI-ALT-05-10 for the Implementation of BWRVIP 38 and -76 in Lieu of B-N-2 Examinations for Core Support Structures NL-19-1336, Implementation of BWRVIP-38 and -76 in Lieu of B-N-2 Examinations for Core Support Structures2019-11-0404 November 2019 Implementation of BWRVIP-38 and -76 in Lieu of B-N-2 Examinations for Core Support Structures ML18320A0572018-11-30030 November 2018 Plant, Units 1 and 2 - Proposed Inservice Inspection Alternative GEN-ISI-ALT-2017-03, Code Case N-513-4 for Moderate Pressure NL-18-0713, Request for Alternative HNP-ISI-ALT-05-08 to Implement Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping for Hatch Unit 1 Plant Service Water2018-05-17017 May 2018 Request for Alternative HNP-ISI-ALT-05-08 to Implement Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping for Hatch Unit 1 Plant Service Water NL-18-0408, Proposed Alternative RR-V-12 Regarding Main Steam Safety Relief Valve Testing2018-04-0909 April 2018 Proposed Alternative RR-V-12 Regarding Main Steam Safety Relief Valve Testing NL-18-0428, Withdrawal of Proposed Alternative GEN-ISI-ALT-2017-001 Implementation of Code Case N-786-12018-04-0909 April 2018 Withdrawal of Proposed Alternative GEN-ISI-ALT-2017-001 Implementation of Code Case N-786-1 ML17279A0452017-10-26026 October 2017 Relief Requests ISI-RR-13, -14, -18, -19, -23, and -24 for Relief from Inservice Inspection Requirements ML17268A0442017-10-20020 October 2017 Relief Requests ISI RR 16, ISI-RR-17, ISI-44-21, and ISI-RR-22 for Relief from Inservice Inspection Requirements ML17205A3452017-08-10010 August 2017 Relief Request HNP-ISI-RR- 05-01 Regarding Reactor Pressure Vessel Head Stud Inservice Inspection Requirements ML17062A8322017-03-29029 March 2017 Relief Request ISI RR-15 Regarding Control Rod Drive Housing Welds Inservice Inspection Requirements ML16314A1322016-11-23023 November 2016 Request for Alternative HNP-ISI-ALT-05-03, Version 1.0, Regarding Reactor Pressure Vessel Flange Leak-Off Piping ML15310A4062015-12-30030 December 2015 Inservice Testing Program Relief Requests and Alternatives for Pumps and Valves - Fifth Ten-Year Interval ML15352A2942015-12-28028 December 2015 Relief from the Requirements of the ASME Code ML15349A9732015-12-18018 December 2015 Relief from the Requirements of the ASME Code NL-15-1221, Withdrawal of RR-V-2 and Response to Request for Additional Information on RR-V-12015-07-16016 July 2015 Withdrawal of RR-V-2 and Response to Request for Additional Information on RR-V-1 NL-15-1265, E.I Hatch, Units 1 and 2 - 10CFR 50.55a Request No. HNP-ISI-ALT-5-01, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Maintaining Isi/Cii Related Activities on the 2001 E/2003A ASME Section XI Code2015-07-16016 July 2015 E.I Hatch, Units 1 and 2 - 10CFR 50.55a Request No. HNP-ISI-ALT-5-01, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Maintaining Isi/Cii Related Activities on the 2001 E/2003A ASME Section XI Code NL-15-1096, Provides Follow-up Letter Regarding Submittal of the Inservice Testing Program Relief Requests and Alternatives for Pumps and Valves - Fifth Ten-Year Interval2015-06-18018 June 2015 Provides Follow-up Letter Regarding Submittal of the Inservice Testing Program Relief Requests and Alternatives for Pumps and Valves - Fifth Ten-Year Interval NL-11-0162, Submittal of Revision to Relief Request RR-V-4 Fourth 10-Year Interval Inservice Testing Program2011-01-26026 January 2011 Submittal of Revision to Relief Request RR-V-4 Fourth 10-Year Interval Inservice Testing Program ML1100504942011-01-14014 January 2011 Safety Evaluation of Relief Request HNP-ISI-ALT-10, Version 1, for the Fourth 10-Year Inservice Inspection Interval Temporary Non-Code Repair of Service Water Piping, TAC ME4253 ML1020304442010-07-22022 July 2010 E-mail from Robert Martin, Regarding Verbal Authorixation of Relief - Note to File ML1020304522010-07-21021 July 2010 Verbal Authorization for Relief Request HNP-ISI-ALT-10 Temporary Non-code Repair of Service Water Piping ML1009802142010-04-0808 April 2010 Safety Evaluation of Relief Request HNP-ISI-ALT-09, Version 2.0, for the Fourth 10-year Inservice Inspection Interval ML0731301882007-12-0606 December 2007 Safety Evaluation for Alternative ISI-ALT-08 ML0713602972007-06-0505 June 2007 Relief, Evaluation of Third 10-Year Interval Inservice Inspection Program Plan Request to Relief Nos. RR-42 Through RR-45, RR- 51, RR-58, RR-59, RR-60, & RR-62 ML0604502862006-02-14014 February 2006 Request for Relief from the Requirements of the American Society of Mechanical Engineered Boiler and Vessel Code (ASME Code) ML0534700912006-01-0303 January 2006 Relief, Boiler and Pressure Vessel Code, MC6528 and MC6529 ML0533303392005-12-22022 December 2005 Request for Relief from the Requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) ML0529700082005-11-0909 November 2005 Relief, Proposed Alternative RR-41 for Third and Fourth 10-Year Inservice Inspection Interval NL-05-1372, Third and Fourth 10-Year Interval Inservice Inspection Programs, Submittal of Revised Exemption Request2005-08-0202 August 2005 Third and Fourth 10-Year Interval Inservice Inspection Programs, Submittal of Revised Exemption Request ML0510301992005-04-25025 April 2005 RR, ISI Program Intervals for Alternative Alignment of Iwe/Iwl Inspection Program (Tac No. MC4870, MC4871, MC4872, MC4873, MC4874, MC4875) NL-05-0726, Fourth 10-Year Interval IST Program Update2005-04-20020 April 2005 Fourth 10-Year Interval IST Program Update ML0501303172005-01-28028 January 2005 Ltr, RR No. 38 NL-04-1764, Third 10-Year Interval Inservice Inspection Program Submittal of Revised Relief Request RR-382004-09-13013 September 2004 Third 10-Year Interval Inservice Inspection Program Submittal of Revised Relief Request RR-38 ML0332800372003-11-21021 November 2003 Safety Evaluation Re. Request to Use ASME Code Case N-661 NL-03-1744, Third 10-Year Interval Inservice Testing Program Submittal of Revised Relief Request RR-V-18 and Response to Request for Additional Information (RAI)2003-09-12012 September 2003 Third 10-Year Interval Inservice Testing Program Submittal of Revised Relief Request RR-V-18 and Response to Request for Additional Information (RAI) ML0304100732003-02-10010 February 2003 Relief Request, Third 10-year Inservice Inspection Program ML0301404462003-01-14014 January 2003 Relief Request, Third 10-Year Inservice Inspection Program ML0230903232002-10-30030 October 2002 Third 10-Year Interval Inservice Testing Program, Revision to Existing Relief Request RR-V-11 ML0218305772002-07-0202 July 2002 Relief Requests for the Second 10-Year Inservice Inspection (ISI) Interval 2020-06-03
[Table view] Category:Letter
MONTHYEARIR 05000321/20230042024-01-31031 January 2024 Integrated Inspection Report 05000321/2023004 and 05000366/2023004 NL-24-0014, Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical.2024-01-30030 January 2024 Revised Response to Request for Additional Information Regarding License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical. ML24012A0652024-01-30030 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) IR 05000321/20233012024-01-17017 January 2024 NRC Operator License Examination Report 05000321/2023301 and 05000366/2023301 ML23341A2042024-01-12012 January 2024 Request for Additional Information Exemption Requests for Physical Barriers (EPID L-2023-LLE-0018 & L-2023-LLE-0021) ML23345A1312024-01-0303 January 2024 Withholding Letter - SNC Fleet - Physical Barriers Exemption (L-2023-LLE-0018 and L-2023-LLE-0021) NL-23-0889, Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water (RHRSW) and Plant Service Water (Psw) Systems2023-12-0606 December 2023 Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water (RHRSW) and Plant Service Water (Psw) Systems IR 05000321/20234022023-11-29029 November 2023 Security Baseline Inspection Report 05000321/2023402, 05000366/2023402, and 07200036/2023401 NL-23-0879, Request for Exemption from Security Event Notification Implementation2023-11-29029 November 2023 Request for Exemption from Security Event Notification Implementation NL-23-0841, Update to Notice of Intent to Pursue Subsequent License Renewal2023-11-20020 November 2023 Update to Notice of Intent to Pursue Subsequent License Renewal IR 05000321/20230112023-11-17017 November 2023 Biennial Problem Identification and Resolution Inspection Report 050003212023011 and 050003662023011 ML23324A4472023-11-14014 November 2023 302 Exam Approval Letter IR 05000321/20230032023-11-0202 November 2023 Integrated Inspection Report 05000321/2023003, and 05000366/2023003 IR 05000321/20234032023-10-26026 October 2023 Cyber Security Inspection Report 05000321/2023403 and 05000366/2023403 IR 05000321/20234012023-10-17017 October 2023 Security Baseline Inspection Report 05000321 2023401 and 05000366 2023401 ML23250A0472023-09-20020 September 2023 Request for Additional Information Regarding License Amendment Request to Relax Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts in Technical Specifications Table 1.1-1, Modes ML23241B0212023-09-12012 September 2023 Review of Quality Assurance Topical Report IR 05000321/20230052023-08-27027 August 2023 Updated Inspection Plan for Edwin I. Hatch Nuclear Plant, Units 1 & 2 - Report 05000321/2023005 and 05000366/2023005 NL-23-0658, Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal2023-08-11011 August 2023 Southern Nuclear Operating Company - Response to Request for Additional Information Regarding Quality Assurance Topical Report Submittal NL-23-0542, CFR 50.46 ECCS Evaluation Model Annual Report for 20222023-08-0909 August 2023 CFR 50.46 ECCS Evaluation Model Annual Report for 2022 IR 05000321/20230022023-08-0808 August 2023 Integrated Inspection Report 07200036/2023001, 05000321/2023002, and 05000366/2023002 NL-23-0624, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2023-08-0404 August 2023 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis ML23202A1122023-08-0202 August 2023 Units, 1 and 2; and Vogtle Units 1 and 2 - Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML23201A0292023-07-20020 July 2023 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection NL-23-0566, ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use2023-07-13013 July 2023 ISFSI and Edwin I. Hatch Nuclear Plant ISFSI - Registration of Spent Fuel Cask Use NL-23-0555, Request for Exemption from Physical Barrier Requirement2023-07-0707 July 2023 Request for Exemption from Physical Barrier Requirement ML23178A0012023-06-27027 June 2023 Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000321 2023403 and 05000366 2023403 NL-23-0444, Quality Assurance Topical Report Submittal2023-06-15015 June 2023 Quality Assurance Topical Report Submittal NL-23-0457, ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use2023-06-12012 June 2023 ISFSI, and Edwin I. Hatch Nuclear Plant, ISFSI - Registration of Spent Fuel Cask Use IR 05000321/20230102023-06-0909 June 2023 Focused Engineering Inspection Report 05000321 2023010 and 05000366 2023010 NL-23-0422, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 2R272023-05-30030 May 2023 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 2R27 NL-23-0383, SNC Response to Regulatory Issue Summary 2023-01:Preparation And.2023-05-19019 May 2023 SNC Response to Regulatory Issue Summary 2023-01:Preparation And. ML23136B2942023-05-19019 May 2023 Summary of May 02, 2023, Observation Public Meeting Held with Southern Nuclear Operating Company, Inc., Regarding a LAR for the Reactor Pressure Vessel Head Closure Bolts for Edwin I. Hatch Nuclear Plant, Units 1 and 2 NL-23-0372, Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 20222023-05-10010 May 2023 Units 1 & 2, Joseph M. Farley Nuclear Plant - Units 1 & 2, Vogtle Electric Generating Plant - Units 1 & 2, Annual Radiological Environmental Operating Reports for 2022 ML23123A2222023-05-0101 May 2023 Notification of Licensed Operator Initial Examination 05000321/2023301, and 05000366/2023301 IR 05000321/20230012023-04-26026 April 2023 Integrated Inspection Report 05000321/2023001 and 05000366/2023001 NL-23-0310, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20222023-04-25025 April 2023 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2022 NL-23-0019, GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2023-04-12012 April 2023 GEN-ISI-ALT-2023-01, Request to Use Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI NL-23-0263, Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report2023-04-0505 April 2023 Southern Nuclear Company Submittal of Drug and Alcohol Testing Errors Identified 10 CFR 26.719(c) 30-Day Report NL-23-0014, Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding2023-03-29029 March 2023 Southern Nuclear Operating Co Submittal of Report on Status of Decommissioning Funding NL-23-0208, Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update2023-03-29029 March 2023 Independent Spent Fuel Storage Installation ISFSI, Decommissioning Funding Plan Triennial Update NL-23-0228, Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21)2023-03-20020 March 2023 Nuclear Property Insurance Coverage as of April 1, 2023 and Licensee Guarantees of Payment of Deferred Premiums (10 CFR 140.21) NL-23-0179, National Pollutant Discharge Elimination System (NPDES) Permit Renewal2023-03-13013 March 2023 National Pollutant Discharge Elimination System (NPDES) Permit Renewal ML23065A1092023-03-0808 March 2023 Correction of Relief Request HNP-ISI-RR-05-02 Allowing to Continue Operation with a Flaw Indication on One Reactor Pressure Vessel Closure Bolt Until December 31, 2025 ML23053A2182023-03-0303 March 2023 Correction of Amendment Nos. 321 and 266 Regarding License Amendment Request to Revise the (National Fire Protection Association) NFPA-805 Fire Protection Program IR 05000321/20220062023-03-0101 March 2023 Annual Assessment Letter Edwin I. Hatch Nuclear Plant, Units 1 and 2 - NRC Inspection Report 05000321/2022006 and 05000366/2022006 NL-23-0101, Cycle 28 Core Operating Limits Report Version 12023-02-23023 February 2023 Cycle 28 Core Operating Limits Report Version 1 ML23034A2462023-02-21021 February 2023 Audit Summary for License Amendment Request to Relax Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs ML22363A3932023-02-15015 February 2023 Issuance of Amendments Nos. 321 and 266, Regarding License Amendment Request to Revise the (National Fire Protection Association) NFPA-805 Fire Protection Program ML22346A1482023-02-0909 February 2023 Issuance of Amendments 320 and 265, Regarding LAR to Revise Technical Specifications to Adopt TSTF-208 and Administrative Correction for Duplicate Technical Specifications 3.4.10 2024-01-31
[Table view] Category:Safety Evaluation
MONTHYEARML23202A1122023-08-0202 August 2023 Units, 1 and 2; and Vogtle Units 1 and 2 - Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML23034A2462023-02-21021 February 2023 Audit Summary for License Amendment Request to Relax Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs ML22363A3932023-02-15015 February 2023 Issuance of Amendments Nos. 321 and 266, Regarding License Amendment Request to Revise the (National Fire Protection Association) NFPA-805 Fire Protection Program ML22346A1482023-02-0909 February 2023 Issuance of Amendments 320 and 265, Regarding LAR to Revise Technical Specifications to Adopt TSTF-208 and Administrative Correction for Duplicate Technical Specifications 3.4.10 ML22297A1462022-12-22022 December 2022 Issuance of Amendments Nos. 319 and 264, Regarding Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF 505, Revision 2, Provide Risk-Informed Extended Completion Times RITS ML22293A0302022-11-0707 November 2022 Amendments 318 & 263 Issuance, Regarding LAR to Adopt TSTF-227, Revision to EOC-RPT Pump Actions, and TSTF-297, ATWS-RPT (EPID L-2021-LLA- 0227) ML22192A1172022-09-0202 September 2022 Issuance of Amendments Nos. 317 & 262, Regarding License Amendment Request to Revise Technical Specifications to Adopt TSTF-207-A, Completion Time for Restoration of Various Excessive Leakage Rates ML22192A1992022-08-19019 August 2022 Issuance of Amendments Nos. 316 and 261, Regarding Request to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling ML22101A0942022-05-20020 May 2022 Issuance of Amendments Nos. 315 and 260, Regarding Request to Eliminate Automatic Main Steam Line Isolation on High Turbine Building Area Temperature ML21349A5182022-02-0202 February 2022 Issuance of Amendments Regarding Revision to Technical Specifications to Adopt TSTF-554, Revise RCS Leakage ML21270A0862021-11-18018 November 2021 Issuance of Amendments Regarding Revision to Technical Specifications 5.7, High Radiation, Administrative Controls ML21264A6442021-09-24024 September 2021 Issuance of Amendment to Revise Technical Specification 3.7.2, Plant Service Water (Psw) System and Ultimate Heat Sink (UHS) (EPID L-2021-LLA-0164) (Emergency Circumstances) ML21217A0912021-09-21021 September 2021 SNC Fleet, Issuance of Amendments Regarding Revision to Emergency Plan to Change Staffing and Extend Staff Augmentation Times for Emergency Response Organization Positions ML21167A3152021-07-30030 July 2021 Issuance of Amendments Regarding Revision to Technical Specifications to Adopt TSTF-541, Rev.2, Add Exceptions to Surveillance Requirements for Valves & Dampers Locked in Actuated Position ML21161A2012021-06-22022 June 2021 Associated Independent Spent Fuel Storage Facilities - Reduction in Commitment to the Quality ML21109A3592021-04-22022 April 2021 Issuance of Amendment No. 254, Regarding Revision to Technical Specifications for One-Time Extension of Completion Times Related to Residual Heat Removal System (Emergency Circumstances) ML20294A0762020-12-0707 December 2020 Issuance of Amendments Nos. 308 and 253, Regarding Application to Revise Technical Specifications to Adopt TSTF-582, Reactor Pressure Vessel Water Inventory Control Enhancements ML20254A0752020-09-18018 September 2020 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 307 to Facility Operating License No. DPR-57 and Amendment No. 252 to Facility Operating License No. NPF-5 ML20202A0052020-09-15015 September 2020 Issuance of Amendments Nos. 306 and 251, Regarding the Application to Revise Technical Specifications to Adopt TSTF-568, Revise Appliciability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2 ML20077J7042020-06-26026 June 2020 Issuance of Amendment Nos. 305 and 250, Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors ML20066F5922020-06-11011 June 2020 Issuance of Amendments Nos. 304 and 249, Regarding License Amendment Request to Adopt NFPA-805 Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) ML20143A2332020-06-0303 June 2020 Relief Request for Deferral of ASME Boiler and Pressure Vessel Code, Section XI, Required Summary Report Submittal (EPID L-2020-LLR-0074 (COVID-19)) ML20034E8942020-02-11011 February 2020 Relief Request Gen-ISI-ALT-2019-01 for Proposed Alternative to Use of ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 ML19337C3222020-01-29029 January 2020 Issuance of Amendments Revision to Technical Specifications to Adopt TSTF-563 ML19239A2442019-11-19019 November 2019 Issuance of Amendments Regarding Revision to Technical Specification Requirements During Handling Irradiated Fuel and Core Alterations ML19267A0232019-11-13013 November 2019 Issuance of Amendments the Revision to Technical Specifications to Adopt TSTF-566, Revise Actions for Inoperable (Residual Heat Removal) RHR Shutdown Cooling Subsystems, Using Consolidated Line Item Improvement Process ML19212A0542019-09-20020 September 2019 Issuance of Amendments 299 and 244 Regarding Revision to Technical Specifications to Adopt Tstf-564, Safety Limit (Minimum Critical Power Ratio) MCPR ML19198A1042019-09-0404 September 2019 Issuance of Amendments Regarding Revision to Technical Specification 3.6.4.1 - Secondary Containment ML19177A1662019-07-0808 July 2019 Edwin Hatch Unit Nos. 1 and 2, Issuance of Amendments Regarding License Amendment Request to Correct Non-Conservative Technical Specification Allowable Values for the Condensate Storage Tank Low Level Transfer Function ML19091A2912019-04-30030 April 2019 Issuance of Amendments 296 and 241 to Revise Technical Specifications 3.6.2.5, Condition C, Residual Heat Removal Drywell Spray End State ML19064A7742019-04-26026 April 2019 Issuance of Amendments Regarding License Amendment Request to Revise Technical Specification 5.2.2.g and Update Emergency Plan Minimum on Shift Staff Tables ML19044A5552019-04-0303 April 2019 Safety Evaluation Regarding Implementation of Hardened Containment Vents Capable of Operation Under Severe Accident Conditions Related to Order EA-13-109 ML19064A0872019-03-0808 March 2019 Proposed Inservice Inspection Alternative HNP-ISI-ALT-05-08 ML19053A0932019-02-22022 February 2019 Issuance of Amendments to Revise Technical Specification 3.8.1, AC Sources - Operating (EPID L-2019-LLA-0026) (Emergency Circumstances) ML19035A5502019-02-0606 February 2019 Relief Request HNP ISI RR 05-02 Regarding Reactor Pressure Vessel Head Stud Inservice Inspection Requirements ML19010A0092019-01-28028 January 2019 Issuance of Amendments to Revise Technical Specification 3.3.8.1, Loss of Power (LOP) Instrumentation ML19011A0102019-01-22022 January 2019 Proposed Alternative HNP-ISI-ALT-05-04 for the Implementation of BWRVIP Guidelines ML18320A0572018-11-30030 November 2018 Plant, Units 1 and 2 - Proposed Inservice Inspection Alternative GEN-ISI-ALT-2017-03, Code Case N-513-4 for Moderate Pressure ML18289A6192018-10-18018 October 2018 Proposed Inservice Inspection Alternative HNP-ISI-ALT-05-07 ML18222A2962018-08-29029 August 2018 Issuance of Amendments to Revise Technical Specification 3.7.7, ECCS - Operating (CAC Nos. MF9997 and MF9998; EPID L-2017-LLA-0263) ML18183A0732018-07-26026 July 2018 Unites 1, 2, 3, and 4; and Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2(CAC Nos. MG0188, MG0189, MG0191, MG0192, MG0193, MG0194, and MG0195; EPID L-2017-LLA-0293) ML18163A1602018-06-15015 June 2018 Proposed Alternative RR-V-11, Extension of Main Steam Safety Relief Valve Test Interval ML18123A3682018-05-31031 May 2018 Issuance of Amendments to Adopt TSTF-542, RPV Water Inventory Control (CAC Nos. MF9662 and MF9663; EPID L-2017-LLA-0215) ML18102B0162018-04-16016 April 2018 Proposed Inservice Inspection Alternative HNP-ISI-ALT-05-05 (CAC Nos. MF9812 and MF9813; EPID L-2017-LLR-0053) ML18022A1072018-01-25025 January 2018 Inservice Inspection Alternative GEN-ISI-ALT-2017-02, Version 1.0, Regarding the Use of ASME Code Case N-789-1 (CAC Nos. MF9942, MF9943, MF9944, and MF9945; EPID L-2017-LLR-0059) ML17355A4402018-01-22022 January 2018 Issuance of Amendments to Revise TS 3.6.4.1, Secondary Containment (CAC Nos. MF9590 and MF9591; EPID L-2017-LLA-0216) ML17271A3072017-11-30030 November 2017 Issuance of Amendments to Revise Actions of TS 5.5.12, Primary Containment Leakage Rate Testing Program ML17279A0452017-10-26026 October 2017 Relief Requests ISI-RR-13, -14, -18, -19, -23, and -24 for Relief from Inservice Inspection Requirements ML17268A0442017-10-20020 October 2017 Relief Requests ISI RR 16, ISI-RR-17, ISI-44-21, and ISI-RR-22 for Relief from Inservice Inspection Requirements ML17208A2312017-08-29029 August 2017 Issuance of Amendments Regarding the Adoption of TSTF 500, DC Electrical Rewrite - Update to TSTF 360 2023-08-02
[Table view] |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 8, 2010 Mr. Mark J. Ajluni Manager, Nuclear Licensing Southern Nuclear Operating Company, Inc 40 Inverness Center Parkway Birmingham, Alabama 35201
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2, SAFETY EVALUATION OF RELIEF REQUEST Ht\lP-ISI-ALT-09, VERSION 2.0, FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NOS. ME3327 AND ME3328)
Dear Mr. Ajluni:
By letter to the U.S. Nuclear RegUlatory Commission (NRC), dated February 16, 2010, as supplemented March 29, 2010 (References 1 and 2, respectively), Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted request for relief HNP-ISI-ALT-09 from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at the Edwin I. Hatch Nuclear Plant, Unit 2. Specifically, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii), the licensee proposed an alternative to the pressure test requirements of ASME Code,Section XI, IWA-4540(c), for repair and replacement activities for mechanical joints made in the installation of pressure-retaining items. The licensee requested implementation of this alternative for activities during a mid-cycle maintenance outage scheduled for April 2010, during the fourth 1O-year interval.
Based on the review of the information the licensee provided, the NRC staff concludes that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety and the alternatives provide reasonable assurance of structural integrity. Therefore, the licensee's proposed alternative is authorized in accordance with 10 CFR 10 50.55a(a)(3)(ii) for the licensee's fourth 1O-year inservice inspection interval. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Sincerely, Gloria Kulesa, Branch Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366
Enclosure:
Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST HNP-ISI-ALT-09 REGARDING PRESSURE TESTING OF MECHANICAL JOINTS SOUTHERN NUCLEAR OPERATING COMPANY, INC.
EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NUMBERS 50-321 AND 50-366
1.0 INTRODUCTION
By letter dated February 16, 2010 (Reference 1), Southern Nuclear Operating Company (SNC, the licensee), submitted Relief Request (RR) HNP-ISI-ALT-09 for Edwin I. Hatch, Unit 2 (HNP2) to use an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. By letter dated March 29,2010 (Reference 2), SNC revised RR HNP-ISI-ALT-09 as Version 2 (HNP-ISI-ALT-09, Version 2) in response to the U.S. Nuclear Regulatory Commission's (NRC) request for additional information dated March 23,2010. Specifically, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii), the licensee proposed an alternative to the pressure test requirements of ASME Code,Section XI, IWA-4540(c), for repair/replacement activities of mechanical joints made in the installation of pressure-retaining items. The licensee requested implementation of this alternative for activities during a mid-cycle maintenance outage scheduled for April 2010 during the fourth 1O-year interval.
2.0 REGULATORY EVALUATION
NRC regulations in 10 CFR 50.55a(g) specify that inservice inspection (lSI) of nuclear power plant components shall be performed in accordance with the requirements of the ASME Code,Section XI, except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(g)(6)(i) states that the NRC may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest, given the consideration of the burden upon the licensee. 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (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. 10 CFR 50.55a(g)(5)(iii) states that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the NRC and submit, as specified in §50.4, information to support the determinations.
Enclosure
-2 Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2,3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements of 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 limitations and modifications listed in paragraph 50.55a(b)(2). The applicable code of record for the fourth lSI interval for HNP2 is the ASME Code,Section XI, 2001 Edition through the 2003 Addenda. The proposed alternative is sought for the April 2010, maintenance shutdown which is in the fourth 10-year lSI interval which ends on December 31, 2015.
Paragraph 55a(b)(2)(xxvi), Pressure Testing Class 1, 2 and 3 Mechanical Joints, places a limitation on licensees using the ASME Code,Section XI, 2001 Edition and later editions and addenda. This limitation requires that the repair replacement activity provisions of ASME Code,Section XI, IWA-4540(c) of the 1998 Edition of Section XI for pressure testing Class 1,2, and 3 mechanical joints be applied. ASME Code,Section XI, IWA-4540(c), of the 1998 Edition requires that mechanical joints made in the installation of pressure-retaining items shall be pressure tested in accordance with ASME Code,Section XI, IWA-5211(a).
ASME Code,Section XI, IWA-5211(a), provides the description of a system leakage test to be done while the system is in operation, during a system operability test, or while the system is at test conditions using an external pressurization source. The test conditions are further described in ASME Code,Section XI, IWB-5221(a), as being conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Request for Alternative ASME Code Components Affected Class 1 pressure retaining mechanical joints.
ASME Code Requirement Affected The ASME Code,Section XI, subarticle IWB-5221 (a) requirement that a Class 1 system leakage test be conducted at a pressure not less than the pressure corresponding to 100 percent of rated reactor power.
Proposed Alternative The licensee will perform a visual examination for leakage (VT-2) of any Class 1 mechanical joints, installed as repair/replacement activities during the April 2010, maintenance shutdown, at an reactor coolant system (RCS) pressure of greater than or equal to 920 psig during reactor startup at approximately 5 percent reactor power versus 1045 psig which is the pressure associated with 100 percent rated reactor power.
-3 Basis for Hardship HNP2 has scheduled a mid-cycle maintenance shutdown for April 2010. The activities scheduled to take place during this maintenance outage include replacement of two safety relief valve (SRV) pilots and an SRV main body and its associated pilot assembly.
These activities would make it necessary to perform the system leakage test required by the provisions of ASME Code,Section XI, IWA-4540(c), in accordance with Paragraph 50.55a(b)(2)(xxvi). The licensee believes performing the system leakage test at 1045 psig would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety over performing the system leakage test at the proposed pressure of greater than or equal to 920 psig.
In References 1 and 2, the licensee describes the hardships related with performing the test at 1045 psig. These hardships would include personnel safety hardships related to the elevated dry well temperatures caused by the additional 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> at RCS temperature required to achieve the pressure increase from 920 psig to 1045 psig.
During these 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, the ambient temperature in the drywell could be expected to increase 13 degrees. These elevated temperatures would cause personnel hazards associated with heat stress as well as burn dangers from elevated component temperatures. The 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to increase pressure from 920 psig to 1045 psig are required by the following operational activities and limitations:
- Control Rod Drive withdrawal limitations and the associated gradual increases in reactor power, pressure and temperature.
- Technical Specification required Pressure versus Temperature limitations.
- Small increases in pressure over time to provide better seating performance of the SRV's Performance of a cold leakage test (non-nuclear heat-up), such as that done following a refueling outage to satisfy the ASME Code,Section XI, IWB-5221, requirements is not desirable following a maintenance outage for the following reasons:
- Extensive valve manipulations, system lineups, and procedural controls are required in order to heat up and pressurize the RCS to establish the necessary test pressure.
- The additional valve lineups and system reconfigurations necessary to support the test would impose an additional challenge to the affected systems.
- These additional valve lineups and system manipulations would add extra radiation exposure to plant personnel.
- Performing a cold leakage test would add approximately two days to the shutdown duration.
-4 Additionally, any leakage discovered during the 920 psig leakage test will be adjusted to account for the marginal increase in leakage rates that might occur at the higher 1045 psig pressure associated with 100 percent rated reactor power. This will ensure the leakage is properly dispositioned in accordance with ASME Code,Section XI, requirements.
The licensee states that drywell monitoring systems are available that would detect leakage that might occur in the mechanical joint connections at higher pressures associated with nominal reactor operation. These systems include drywell air temperature and pressure monitoring and the drywell floor and equipment drain sumps.
3.2 NRC Staff Evaluation ASME Code,Section XI, IWA-4540, provides the pressure testing requirements for repair/replacement activities of Class 1, 2 and 3 items. The code of record for HNP2 specifies leakage or hydrostatic testing is required for repair/replacement activities performed by welding or brazing on a pressure-retaining boundary. This would not require mechanical joints to be leakage tested. However, paragraph 55a(b)(2)(xxvi) was incorporated into the regulations when incorporating the 2001 Edition through the 2003 Addenda of ASME Section XI because the NRC staff found that ASME Code pressure testing of mechanical joints after repair and replacement activities is still warranted.
In accordance with paragraph 50.55a(b)(2)(xxvi), ASME Code,Section XI, IWA-4540(c), of the 1998 Edition of Section XI, requires that mechanical joints made in the installation of pressure retaining items be pressure tested during a system leakage test at nominal operating pressure.
In the 1998 Edition of ASME Code,Section XI, IWB-5221 (a) the test conditions are further described as being conducted at a pressure not less than nominal operating pressure associated with normal system operation. The 2001 Edition through the 2003 Addenda of ASME Code,Section XI, IWB-5221(a), describes a system leakage test as being conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power (i.e. 1045 psig).
In the Federal Register Notice (FRN)(Reference 3) which incorporated paragraph 50.55a(b)(2)(xxvi), the NRC staff explained the reasoning for this requirement. The requirements to pressure test Class 1, 2 and 3 mechanical joints undergoing repair and replacement activities were deleted in the 1999 Addenda of Section XI. There was no justification for eliminating the requirements for pressure testing Class 1, 2 and 3 mechanical joints and the NRC staff believed the pressure testing was necessary to ensure and verify the integrity of the pressure boundary. The FRN never discussed specific pressures of the test to be performed but focused on the need to perform a post repair and replacement pressure test and VT-2 examination to verify the integrity of the pressure boundary. The NRC staff believes the licensee's proposed alternative will verify the leak tightness and structural integrity of the mechanical joints involved.
As an alternative to the system pressure test at operating pressure (1045 psig), the licensee proposed to perform the system leakage test at a pressure of at least 920 psig at approximately 5 percent reactor power. The NRC staff believes that the proposed test pressure
-5 (approximately 90 percent of the pressure corresponding to 100 percent rated reactor power) will be sufficiently high to cause leakage from any mechanical connections following opening and reassembly of the item if the leak-tight connection has not been established. Leakage through the mechanical joint would be detectable at 920 psig with a slightly lower leakage rate than that at 1045 psig which is the ASME Code-required test pressure for the system leakage test. The licensee stated that disposition of any observed leakage at 920 psig will account for the marginal increase in leakage rates that might occur at 1045 psig.
The NRC staff agrees that performance of the VT-2 examination at the ASME Code-required pressure of 1045 psig low reactor power levels versus 920 psig would involve hardship from a personnel safety standpoint. The environmental conditions would require consideration of serious heat stress and valid burn hazard concerns. These conditions would also require additional special safety precautions such as ice vests and cool air supply lines. These adverse conditions and the additional burden of the safety precautions could impact the quality of the leakage examination due to the hardship imposed on the examination personnel. Performing the leakage test at 1045 psig during low power operations would also present operational challenges such as altering normal stearn pressure controls, possible SRV seating issues and Control Rod Drive withdrawal limitations.
The NRC staff also agrees that performance of a cold leakage test (non-nuclear heatup), such as that required following a refueling outage would involve hardship. Performing this type test would require filling the main steam lines and the reactor vessel solid with water. In order to establish the necessary test conditions would require performing extensive temporary hanger modifications, valve lineups and system manipulations. All of these activities would require personnel radiation exposure in addition to normal startup activities.
Based on the above evaluation, the NRC staff finds that performing the alternative, a visual examination for leakage during a system leakage test at equal to or greater than 920 psig, provides reasonable assurance of the leakage and structural integrity of the items, and that compliance with the ASME Code-specified reqUirements would result in hardship without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff concludes that complying with the ASME Code reqUirement would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Furthermore, based on the above, the NRC staff determines that the proposed alternative described in RR HNP-ISI-ALT-09, Version 2, provides reasonable assurance of structural integrity and leak tightness of the subject components. Therefore, the NRC authorizes the proposed alternative in accordance with 10 CFR 50.55a(a)(3)(ii) for the April 2010 maintenance shutdown in the fourth 1O-year inspection interval at HNP2.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
REFERENCES
- 1. Letter, M.J. Ajluni (SNC) To U.S. Nuclear RegUlatory Commission containing lSI
-6 Program Alternative HNP-ISI-ALT-09, Edwin I. Hatch Nuclear Plant, Units 1&2, February 16, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100480010).
- 2. Letter, M.J. Ajluni (SNC) To U.S. Nuclear Regulatory Commission containing lSI Program Alternative HNP-ISI-ALT-09, Edwin I. Hatch Nuclear Plant, Unit 2, March 29, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100890051).
- 3. 69 Fed Reg 58804, 58820 (October 1,2004) Industry Codes and Standards; Amended Requirements.
Principal Contributor: Keith M. Hoffman, NRRIDCI Date of issuance: April 8, 2010
ML100980214 *memo transmitted SE dated OFFICE NRR/LPL2-1/PM NRRlLPL2-1/LA NRRlCPNB/BC OGC NRRlLPL2-1/BC NAME RMartin SRohrer TLupold Waived GKulesa DATE 04/08/10 04/08/10 04/08/10 ML090480460 04/08/10