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Category:Letter
MONTHYEARML24256A0282024-09-12012 September 2024 2024 Hatch Requal Inspection Corporate Notification Letter NL-23-0930, Application to Revise Technical Specifications to Adopt TSTF-591, Revise Risk Informed Completion Time (RICT) Program2024-09-11011 September 2024 Application to Revise Technical Specifications to Adopt TSTF-591, Revise Risk Informed Completion Time (RICT) Program NL-24-0337, Interim 10 CFR 21.21(a)(2) Report Regarding Operation Technology, Inc., ETAP Software Error in Transient Stability Program2024-09-0909 September 2024 Interim 10 CFR 21.21(a)(2) Report Regarding Operation Technology, Inc., ETAP Software Error in Transient Stability Program ML24249A1362024-09-0404 September 2024 EN 57304 - Westinghouse Electric Company, LLC, Final Report - No Embedded Files. Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21 NL-24-0334, 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc2024-09-0303 September 2024 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc IR 05000321/20240912024-08-27027 August 2024 NRC Investigation Report 2-2023-003 and NOV - NRC Inspection Report 05000321/2024091 and 05000366/2024091 IR 05000321/20240052024-08-26026 August 2024 Updated Inspection Plan for Edwin I. Hatch Nuclear Plant, Units 1 and 2 - Report 05000321/2024005 and 05000366/2024005 NL-24-0313, Application to Revise Technical Specifications Surveillance Requirements to Increase Safety/Relief Valves Setpoint Response to Request for Additional Information2024-08-23023 August 2024 Application to Revise Technical Specifications Surveillance Requirements to Increase Safety/Relief Valves Setpoint Response to Request for Additional Information IR 05000321/20240022024-08-0808 August 2024 Edwin I Hatch Nuclear Plants, Units 1 and 2 – Integrated Inspection Report 05000321-2024002 and 05000366-2024002 NL-24-0276, Post-Audit Supplement to License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components2024-07-26026 July 2024 Post-Audit Supplement to License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components NL-24-0290, Response to Request for Additional Information Related to Request for Specific Exemption2024-07-26026 July 2024 Response to Request for Additional Information Related to Request for Specific Exemption NL-24-0261, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20232024-07-19019 July 2024 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2023 ML24198A1252024-07-16016 July 2024 Edwin I Hatch Nuclear Plant Units 1 - 2 Notification of Conduct of Title 10 of the Code of Federal Regulations 50 NL-24-0260, Inservice Inspection Program Owner’S Activity Report (OAR-1) for Refueling Outage 1R312024-07-0909 July 2024 Inservice Inspection Program Owner’S Activity Report (OAR-1) for Refueling Outage 1R31 NL-24-0237, Incorrectly Installed Temporary Modification Results in Multiple Conditions Prohibited by Plant Technical Specifications2024-07-0303 July 2024 Incorrectly Installed Temporary Modification Results in Multiple Conditions Prohibited by Plant Technical Specifications NL-24-0240, Reactor Core Isolation Cooling (RCIC) System Inoperable Due to Mispositioned Link2024-07-0303 July 2024 Reactor Core Isolation Cooling (RCIC) System Inoperable Due to Mispositioned Link NL-24-0143, Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in2024-06-27027 June 2024 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in NL-24-0239, Response to Apparent Violation in NRC Inspection Report 05000321, 366/2024090: EA-23-1392024-06-17017 June 2024 Response to Apparent Violation in NRC Inspection Report 05000321, 366/2024090: EA-23-139 ML24163A0532024-06-14014 June 2024 Audit Plan - Alternative Seismic Method LAR NL-24-0148, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2024-06-0404 June 2024 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis ML24149A0492024-06-0404 June 2024 SNC Fleet - Regulatory Audit in Support of Review of the License Amendment Request to Revise TS 1.1, Use and Application Definitions, and Add New Technical Specification 5.5.21 and 5.5.17, Online Monitoring Program, NL-24-0202, SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations2024-05-24024 May 2024 SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations IR 05000321/20240902024-05-15015 May 2024 NRC Inspection Report 05000321-2024090 and 05000366-2024090, Investigation Report 2-2023-003; and Apparent Violation NL-24-0191, Annual Radiological Environmental Operating Reports for 20232024-05-10010 May 2024 Annual Radiological Environmental Operating Reports for 2023 NL-24-0166, Manual Reactor Trip Due to Loss of Feedwater2024-05-0909 May 2024 Manual Reactor Trip Due to Loss of Feedwater NL-24-0195, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water.2024-05-0707 May 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times for Residual Heat Removal Service Water. NL-24-0064, Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.52024-05-0303 May 2024 Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5 IR 05000032/20240112024-04-25025 April 2024 Notification of Edwin I. Hatch Nuclear Plant - Comprehensive Engineering Team Inspection (CETI) Baseline Inspection Report 0500032/2024011 and 05000366/2024011 NL-24-0165, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20232024-04-25025 April 2024 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2023 NL-24-0157, Incorrectly Installed Temporary Modification Results in a Condition Prohibited by Plant Technical Specifications2024-04-24024 April 2024 Incorrectly Installed Temporary Modification Results in a Condition Prohibited by Plant Technical Specifications IR 05000321/20240012024-04-22022 April 2024 Integrated Inspection Report 05000321/2024001 and 05000366/2024001 NL-24-0026, Application to Revise Technical Specifications Surveillance Requirements to Increase Safety/Relief Valves Setpoint2024-04-19019 April 2024 Application to Revise Technical Specifications Surveillance Requirements to Increase Safety/Relief Valves Setpoint NL-24-0144, Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)2024-04-0909 April 2024 Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La) NL-24-0115, Response to Request for Additional Information Exemption Requests for Physical.2024-04-0404 April 2024 Response to Request for Additional Information Exemption Requests for Physical. NL-24-0116, Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred.2024-03-29029 March 2024 Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred. NL-24-0062, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2024-03-12012 March 2024 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis NL-24-0089, Correction of Technical Specification Omission2024-03-0909 March 2024 Correction of Technical Specification Omission ML24069A0012024-03-0909 March 2024 – Correction of Amendment No. 266 Regarding License Amendment Request Regarding Relocation of Specific Surveillance Frequencies to a Licensee-Controlled Program (TSTF-425, Revision 3) ML24047A0362024-03-0404 March 2024 Response to Hatch and Vogtle FOF Dates Change Request (2025) NL-24-0061, Cycle 32 Core Operating Limits Report Version 12024-03-0101 March 2024 Cycle 32 Core Operating Limits Report Version 1 IR 05000321/20230062024-02-28028 February 2024 Annual Assessment Letter Edwin I. Hatch Nuclear Plant Units 1 and 2 - Nrc Inspection Report 05000321/2023006 and 05000366/2023006 NL-24-0067, 10 CFR 72.48(d)(2) Biennial Report2024-02-26026 February 2024 10 CFR 72.48(d)(2) Biennial Report NL-24-0051, License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components2024-02-20020 February 2024 License Amendment Request to Revise Renewed Facility Operating Licenses to Adopt an Alternative Seismic Method for Categorization of Structures, Systems, and Components NL-24-0042, Response to Request for Additional Information Exemption Requests for Physical Barriers2024-02-13013 February 2024 Response to Request for Additional Information Exemption Requests for Physical Barriers NL-24-0033, Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers2024-02-0505 February 2024 Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers IR 05000321/20230042024-01-31031 January 2024 Integrated Inspection Report 05000321/2023004 and 05000366/2023004 ML24012A0652024-01-30030 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) 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. 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) 2024-09-09
[Table view] Category:Report
MONTHYEARNL-24-0334, 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc2024-09-0303 September 2024 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc ML24047A2092024-02-22022 February 2024 Calendar Year 2023 Baseline Inspection Completion ML24005A1142024-01-0505 January 2024 Recommendation for 2023-301 Cr/Sim 3 (Emergency Depress the Reactor Using Main Steam Line Drains) NL-22-0510, Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20212022-07-14014 July 2022 Plants Units 1 and 2, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2021 A000412, National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application2021-12-0202 December 2021 National Pollutant Discharge Elimination System (NPDES) Permit Renewal Application NL-20-1295, 10 CFR 71.95 Report on Non-Compliance Involving Radwaste Cask 3-60B2020-12-14014 December 2020 10 CFR 71.95 Report on Non-Compliance Involving Radwaste Cask 3-60B ML20303A1782020-09-29029 September 2020 Submittal of Revision 38 to Updated Final Safety Analysis Report NL-19-0674, Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 12019-09-30030 September 2019 Proposed Alternative to Use ASME Code Case N-831-1, Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic or Austenitic Pipe Section XI, Division 1 NL-18-0863, CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations2018-06-21021 June 2018 CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations NL-18-0506, Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test2018-04-16016 April 2018 Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-18-0282, Enclosure 1: NFPA 805 LAR Transition Report for Edwin I. Hatch2018-04-0404 April 2018 Enclosure 1: NFPA 805 LAR Transition Report for Edwin I. Hatch NL-17-1916, Pressure and Temperature Limits Report2017-11-27027 November 2017 Pressure and Temperature Limits Report ML18012A0582017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33884, Revision 0, Gnf Fecrai Atf Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai Atf Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-10-31031 October 2017 NEDO-33884, Revision 0, Gnf Fecrai ATF Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-18-0026, NEDO-33883, Revision 0, Gnf Armor Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1.2017-09-30030 September 2017 NEDO-33883, Revision 0, Gnf Armor Lead Test Assembly for Edwin I. Hatch Nuclear Plant, Unit 1. NL-17-1255, Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope High Frequency Confirmation Evaluation2017-08-22022 August 2017 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope High Frequency Confirmation Evaluation NL-17-0955, Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-062017-06-0505 June 2017 Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06 ML17069A2402017-04-13013 April 2017 Mitigating Strategies Assessment (CAC Nos. MF7932 and MF7933) - Redacted ML16356A0172016-12-16016 December 2016 Information Report for Lead Use Assemblies NL-16-2466, Fukushima Near-Term Task Force Recommendation 2.1 Expedited Seismic Evaluation Process Report Completion2016-12-15015 December 2016 Fukushima Near-Term Task Force Recommendation 2.1 Expedited Seismic Evaluation Process Report Completion NL-16-1136, Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope Low Frequency Evaluation2016-08-10010 August 2016 Fukushima Near-Term Task Force Recommendation 2.1 Seismic Limited-Scope Low Frequency Evaluation NL-16-0463, Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report2016-04-0101 April 2016 Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report NL-15-2010, E.I. Hatch - Submits 10 CFR 71.95 Report on Non-Conformance Involving Radwaste Cask 8-120B2015-11-0202 November 2015 E.I. Hatch - Submits 10 CFR 71.95 Report on Non-Conformance Involving Radwaste Cask 8-120B NL-15-1461, Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2015-08-21021 August 2015 Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML15097A4242015-04-27027 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the NTTF Review ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis ML14335A1372015-03-25025 March 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) NL-14-1989, Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.12014-12-30030 December 2014 Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.1 NL-14-1876, Proposed Lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0, Conceptual Design Information Package2014-11-24024 November 2014 Proposed Lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0, Conceptual Design Information Package NL-14-1245, 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample2014-08-22022 August 2014 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample ML14223A7942014-07-31031 July 2014 Enclosure 2 - Non-Proprietary Gnf Report GNF-001N6296-R1-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, Hatch 2 Cycle 24 ML14155A4052014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (TAC MF0234-35) ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML14155A3612014-06-0606 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident NL-14-0343, Seismic Hazard and Screening Report for CEUS Sites2014-03-31031 March 2014 Seismic Hazard and Screening Report for CEUS Sites NL-14-0326, Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 20122014-03-0606 March 2014 Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 2012 ML13364A2022014-02-27027 February 2014 Interim Staff Evaluation Related to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14045A1472014-02-12012 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Edwin I. Hatch Nuclear Plant, Units 1 and 2, TAC Nos.: MF0712 and MF0713 NL-13-1898, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2013-08-30030 August 2013 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML13193A3662013-08-0707 August 2013 Request for Concurrence on the Effects of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 on the Federally-Listed Endangered Species Altamaha Spinymussel NL-13-0214, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events..2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events.. NL-13-0172, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051)2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051) NL-13-0402, GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 20082013-01-31031 January 2013 GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 2008 ML13115A4732013-01-31031 January 2013 GNF-0000-0079-7396NP, Rev. 6, Technical Basis Supporting GNF-Ziron Lead Test Assembly Introduction Into the Hatch Nuclear Plant, March 2008 ML12355A0592012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 ML12355A0562012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 ML12355A0552012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 ML12355A0542012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 2024-09-03
[Table view] Category:Miscellaneous
MONTHYEARNL-24-0334, 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc2024-09-0303 September 2024 0 to the Updated Final Safety Analysis Report, Technical Specifications Bases Changes, Technical Requirements Manual Changes, License Renewal 10 CFR 54 .37(b) Changes, 10 CFR 50.59 Summary Report & Revised Nrc ML24047A2092024-02-22022 February 2024 Calendar Year 2023 Baseline Inspection Completion ML24005A1142024-01-0505 January 2024 Recommendation for 2023-301 Cr/Sim 3 (Emergency Depress the Reactor Using Main Steam Line Drains) NL-18-0863, CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations2018-06-21021 June 2018 CFR 50.55a Request for Alternative HNP-ISI-ALT-05-04 Implementation of BWRVIP Documents in Lieu of Certain 8-N-1 and 8-N-2 Examinations NL-18-0506, Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test2018-04-16016 April 2018 Failure to Observe the Certificate of Compliance Condition of the 8-120B Cask Pre-Shipment Leak Test NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure2018-04-0606 April 2018 Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure NL-17-0955, Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-062017-06-0505 June 2017 Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06 ML17069A2402017-04-13013 April 2017 Mitigating Strategies Assessment (CAC Nos. MF7932 and MF7933) - Redacted ML16356A0172016-12-16016 December 2016 Information Report for Lead Use Assemblies NL-16-0463, Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report2016-04-0101 April 2016 Submittal of 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2015 and Significant Change/Error Report NL-15-1461, Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2015-08-21021 August 2015 Submittal of 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML15097A4242015-04-27027 April 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations Relating to Recommendation 2.1 of the NTTF Review ML15106A3102015-04-17017 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Draft Minimum Shift Staffing Analysis ML14335A1372015-03-25025 March 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) NL-14-1989, Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.12014-12-30030 December 2014 Expedited Seismic Evaluation Process Report - Fukushima Near-Term Task Force Recommendation 2.1 NL-14-1245, 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample2014-08-22022 August 2014 10 CFR 26.719(c) Report: False Negative Results for a Blind Performance Test Sample ML14155A4052014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (TAC MF0234-35) ML14155A3612014-06-0606 June 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident NL-14-0326, Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 20122014-03-0606 March 2014 Units 1 and 2, Recommendation 2.1 Flood Hazard Reevaluation Report, Requested by NRC Letter Dated March 12, 2012 NL-13-1898, 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B2013-08-30030 August 2013 10 CFR 71.95 Report on Potential Issues Involving Radwaste Cask 8-120B ML13193A3662013-08-0707 August 2013 Request for Concurrence on the Effects of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 on the Federally-Listed Endangered Species Altamaha Spinymussel NL-13-0214, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events..2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events.. NL-13-0172, Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051)2013-02-27027 February 2013 Southern Nuclear Operating Company'S Overall Integrated Plan in Response to 3/12/2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (EA-12-051) NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 NL-12-2268, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 112012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 ML12355A0502012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 NL-12-2268, Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 112012-11-26026 November 2012 Edwin I. Hatch, Unit 1, SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 11 ML12355A6332012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 3 of 11 ML12356A4102012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 9 of 11 ML12356A4092012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 ML12356A4082012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 10 of 11 ML12355A6662012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 7 of 11 NL-12-2269, SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 112012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 6 of 11 ML12355A6622012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 8 of 11 ML12355A6372012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 11 ML12355A6352012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 5 of 11 ML12355A6342012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 ML12355A0522012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 4 of 11 ML12355A6312012-11-26026 November 2012 SNCH082-RPT-02, Ver. 1.0, Edwin I. Hatch, Unit 2 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 11 ML12355A0592012-11-26026 November 2012 SNCH082-RPT-01, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 11 of 11 2024-09-03
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H. 1. Sumner, Jr.
Vice President Hatch Project Southern Nuclear Operating Company, Inc. Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 February 20, 2006 Docket Nos.: 50-32 1 50-366 - - COMPANY Energy to Serve Your Worldw NL-06-03 12 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Revort of Facility Changes, Tests.
and Exveriments Safety Evaluation Summaries Ladies and Gentlemen: Enclosed is the 24 month report of facility changes, tests, and experiments safety evaluation summaries in accordance with the requirements of 10 CFR 50,59(d)(2).
This letter contains no NRC commitments.
If you have any questions, please advise.
H. L. Surnner, Jr.
Enclosure:
Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries cc: Southern Nuclear Operating Comvanv Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, General Manager - Plant Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch Enclosure Edwin I. Hatch Nuclear Plant NRC Docket Nos.: 50-321 and 50-366 Operating Licenses:
DPR-57 and NPF-5 Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries Enclosure Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS ABN AC ADS AHU ALARA APLHGR APRM ARI ARM ARTS ASME ATWS ATWS-RPT as-built notice alternating current automatic depressurization system air handling unit as low as reasonably achievable average power linear heat generation rate average power range monitor alternate rod insertion area radiation monitor average power range monitor, rod block monitor, and Technical Specifications American Society of Mechanical Engineers anticipated transient without scram anticipated transient without scram-recirculation pump trip BHD bottom head drain BOP balance of plant BWR boiling water reactor BWROG Boiling Water Reactor Owners Group CFR Code of Federal Regulations COLR Core Operating Limits Report CRD control rod drive CS core spray CST condensate storage tank DAS DBA DBE DC DCB DCR DCS DHR dP data acquisition system design basis accident design basis earthquake direct current double cantilever beam design change request dry cask storage decay heat removal differential pressure ECCS emergency core cooling system ECP electrochemical potential EDG emergency diesel generator EFCV excess flow check valve EFPD effective full power days EFPH effective full power hours Enclosure Page 2 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Chan~es, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS EHC ELI EM1 EOC-RPT EOF EPA ERFDS ETS EQ electrohydraulic control Equipment Location Index electromagnetic interference end of cycle-recirculation pump trip Emergency Operations Facility Environmental Protection Agency Emergency Response Facility Display System Environmental Technical Specifications Environmental Qualification FHA Fire Hazards Analysis FPC fuel pool cooling FSAR Final Safety Analysis Report GE General Electric GL Generic Letter GPC Georgia Power Company HCU hydraulic control unit HNP Hatch Nuclear Plant HPCI high pressure coolant injection HVAC heating, ventilation, and air-conditioning HWC hydrogen water chemistry I&C IE IGSCC ILRT IRM ISFSI IS1 IST LAN LC0 LDS LDCR LLRT LLS LOCA LOSP instnunentation and control inspection and enforcement intergranular stress corrosion cracking integrated leak rate test intermediate range monitor independent spent fuel storage installation inservice inspection inservice testing local area network limiting condition for operation leak detection system license document change request local leak rate test low-low set loss of coolant accident loss of offsite power Enclosure Page 3 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS LPAP low power alarm point LPCI low pressure coolant injection LPM loose-parts monitor LPRM local power range monitor LPSP low power setpoint MCC MCPR MCR MCRECS MDC MG MPC MOV MPL MSIV MS SRV MSL MSLRM MSR motor control center minimum critical power ratio main control room main control room environmental control system minor design change motor generator Multi-Purpose Canister motor-operated valve master parts list main steam isolation valve main steam safety relief valve main steam line main steam line radiation monitor moisture separator reheater NMA noble metals addition NPSH net positive suction head NRC Nuclear Regulatory Commission NSSS nuclear steam supply system ODCM Offsite Dose Calculation Manual OPDRV operations with the potential to drain the reactor vessel OPRM oscillation power range monitor PAM PASS PCIS PCIV P&ID PLC PPC PRB PRNM PSW PSW post accident monitoring post accident sampling system primary containment isolation system primary containment isolation valve piping and instrumentation diagram programmable logic controller plant process computer Plant Review Board power range neutron monitor plant service water plant service water Enclosure Page 4 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS Q A quality assurance RBM RCIC RCPB RCS REA RES RFI RFP RFPT RG RHR RHRSW RMCS RPS RPT RPV RRS RSCS RWCU or RWC RWCS RWE RWM rod block monitor reactor core isolation cooling reactor coolant pressure boundary reactor coolant system Request for Engineering Assistance Request for Engineering Services radio frequency interference reactor feed pump reactor feed pump turbine Regulatory Guide residual heat removal residual heat removal service water reactor manual control system reactor protection system recirculation pump trip reactor pressure vessel reactor recirculation system rod sequence control system reactor water cleanup reactor water cleanup system rod withdrawal error rod worth minimizer SAER Safety Audit and Engineering Review SAT station auxiliary transformer SBGT or standby gas treatment SGTS or SGT SCM stress corrosion monitor SDC setpoint design change SED System Evaluation Document S JAE steam jet air ejector SLMCPR safety limit minimum critical power ratio SNC Southern Nuclear Operating Company SoRA Summary of Required Actions SPDS Safety Parameter Display System SRB Safety Review Board SR Surveillance Requirement SRM source range monitor SRV safety relief valve Enclosure Page 5 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries GLOSSARY ACRONYMS AND ABBREVIATIONS SSAR SSC TBWD TCV THV TIL TIP TLD TM TRM TS TSV safe shutdown analysis report system, structure, or component thrust bearing wear detector turbine control valve torus hardened vent Technical Information Letter traversing incore probe thennoluminescent dosimeter Temporary Modification Technical Requirements Manual Technical Specifications turbine stop valve Version Enclosure Page 6 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries 10 CFR 50.59 SUMMARIES TEMPORARY MODIFICATIONS TTM) APC 1-04-033 This Temporary Modification via control of an Annunciator and Plant Component sheet (APC) is to disable the "ROD DRIFT" alarm function for rod 26-39 and rod 34-39 via installation of a jumper on the Probe Buffer Card.
The FSAR mentions that a drifting rod is "indicated by an alarm and a red light in the MCR. The rod drift condition is also monitored by the process computer." As a result of this APC, the alarm will not come in for the three rods identified.
This TM will allow the drift alarm to function for all other control rods for which the alarm is not disabled. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR. APC 1-04-089 This Temporary Modification via control of an Annunciator and Plant Component sheet (APC) is to disable the "ROD DRIFT" alarm function for rod 26-39, rod 34- 27 and rod 34-39 via installation of a jumper on the Probe Buffer Card. The FSAR mentions that a drifting rod is "indicated by an alarm and a red light in the MCR. The rod drift condition is also monitored by the process computer." As a result of this APC, the alarm will not come in for the three rods identified. This TM will allow the drift alarm to function for all other control rods for which the alarm is not disabled. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR. TM 1-04-02 1, Rev. 0 This activity is for binding closed or "gagging" 1E 1 1-F200B, which is the minimum flow valve for RHRSW pump 1 E 1 1 -COO 1 B. The FSAR states that the design function of this valve is a low-flow bypass. Since the valve will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR. The valve addressed by this TM cannot form any part of a sequence of events which could cause an accident. Hence, no change to this valve could affect the probability of occurrence of an accident.
Enclosure Page 7 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries DESIGN CHANGE REOUESTS (DCR) DCR 90-028. Rev. 0 This DCR provides the design for replacing nine Allis Chalmers safety related starters. Siemens starter pan assemblies will be used as replacements. The replacements are functionally equivalent and testing will be performed to assure the replacements are seismically and environmentally qualified for worst case conditions.
The margin of safety of the Motor Control Centers will not be affected by the starter assembly change out. DCR 98-047, Rev. 0 This DCP will replace RMS-9 trip devices on selected (based on load category) frames of 600V Busses 2A, 2B, 2AA and 2BB with MVT+ trip units. All the busses and their respective loads are nonsafety-related. The function of the 600V switchgear will not change. The breaker trip devices are being replaced to enhance 600V distribution system reliability by reducing spurious breaker trips.
The analysis of trip unit failure presented in the Discussion section of the safety evaluation addresses the potential impact of EMIIRFI conducted and radiated emissions from the new trip units on any safety-related equipment or system. The conclusion is that no safety-related equipment or system would be adversely affected by the installation of these trip units. Therefore, the probability of an accident previously evaluated in the FSAR will not increase.
DCR 99-050, Rev. 0 This DCR provides the design for the replacement of the Safety Parameter Display System (SPDS) and the Emergency Response Facility Display System (ERFDS). The design functions of the existing SPDS will be retained in the new system, although they will be accomplished using different software and microprocessor-based hardware.
The SPDS has no control functions.
It is a monitoring system only. The same information, including calculated values, will be displayed to the plant operators as the existing system. This DCR is considered to be a digital upgrade.
EMYRFI testing has demonstrated that the RTP racks are not expected to adversely impact any safety related or important to safety system. All interconnections between Class 1E equipment and non-class 1 E components for the new SPDS will be properly isolated.
DCP 1H03-009, Rev. 3 This design change will provide adequate margin for the Unit 1 reactor building crane main hoist to lift and lower the rated 125 ton loads, improve crane Enclosure Page 8 of 14 Edwin I. Hatch Nuclear Plant Rwort of Facility Changes, Tests. and Experiments Safety Evaluation Summaries reliability, and bring the crane into compliance with OSHA requirements.
The addition of a pulley on the main hoist drum, which is coupled to an auxiliary shaft via a gear belt, introduces a new potential failure.
This change may be considered adverse. However, there are no design basis accidents in the FSAR for which the crane is the initiator.
The pulley driven overspeed switch provides a backup method for preventing a load drop. The Unit 1 Reactor Building Crane is a single failure proof crane for which a load drop is not considered a credible event. The new switch, belt and pulley assembly is considered to be as rugged as the prior system. Because the potential failures are similar, no new possibility of a malhction of an SSC important to safety with a different result than previously evaluated in the Updated FSAR is expected by this modification.
DCP 1 H03-026T, Rev. 0 This DCR adds a blind plate in place of the 1 P41 -Dl66 orifice plate, located in the 1B diesel generator room, and remove a section of the 12" shield piping in the 2G switchgear room and relocate service water vent and drain valves.
This design change will not affect the service water supply to any component except the 1B diesel generator. This change will remove the Division I back-up supply. This does not reduce the reliability of any component.
It only reduces the redundancy in available back-up cooling water supplies.
PRA evaluation shows that there is no increase in the average risk following implementation of this design change.
MDC 03-5009, Rev.
0 This MDC will permanently remove the automatic PSW transfer and isolation logic for the 1Z41 -B008B MCR AC unit. This is a change to the plant as described in the Unit 1 FSAR section 10.7.6. This logic automatically transfers PSW supply fiom Div. I to Div.
I1 in the event of a low flow condition in Div. I in conjunction with a loss of offsite power or a loss of coolant accident. Therefore, this logic is designed to function following a LOCA or LOSP and the subsequent occurrence of a problem causing a low flow in Div. I of the PSW system. This change has no impact upon the probability of a problem occurring.
The overall system operation will remain as described in the FSAR. Therefore, the probability of a system failure remains unaffected as well. DCP 10401 13801, Ver. 1 This DCP is to implement setpoint and calibration changes to facilitate implementation of the 10 PSI reactor pressure increase to allow the achievement of 100 percent of the rated power approved under Appendix K uprate. Because of a slight increase in rated containment pressure the License Amendments for REA 00-650lRER 2003-254 (RPV 10 PSI Increase, LDCR-2003-077 (Tech Spec Changes) & LDCR 2004-040 (FSAR Revisions) that allows the increase must be approved prior to implementation of this DCP. LDCR 2004-041 which is a result Enclosure Page 9 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Exveriments Safetv Evaluation Summaries of the change in steam density is included with this DCP. This DCP is to implement the setpoint and calibration changes associated with the 10 PSI increase.
There are no impacts to the frequency or occurrence of accidents, likelihood of occurrence of a malfunction of a structure, system, or component (SSC), consequences of accidents previously evaluated, consequences of SSC malfunctions, possibility for the creation of an accident of a different type, possibilities for malfunctions of SSC's, impact on the fuel cladding, reactor coolant pressure boundary, or containment, or changes in the method of evaluations due to the changes associated with this package. Were the 10 PSI increase not to occur and these changes were implemented the result would be a reduction in operating margin rather than safety margin. DCP 2040 1 13901, Ver.
1 This DCP is to implement setpoint and calibration changes to facilitate implementation of the 10 PSI reactor pressure increase to allow the achievement of 100 percent of the rated power approved under Appendix K uprate. Because of a slight increase in rated containment pressure the License Amendments for REA 00-650RER 2003-254 (RPV 10 PSI Increase, LDCR-2003-077 (Tech Spec Changes) & LDCR 2004-040 (FSAR Revisions) that allows the increase must be approved prior to implementation of this DCP. LDCR 2004-039 which is a result of the change in steam density is included with this DCP. This DCP is to implement the setpoint and calibration changes associated with the 10 PSI increase. There are no impacts to the frequency or occurrence of accidents, likelihood of occurrence of a malfunction of a structure, system, or component (SSC), consequences of accidents previously evaluated, consequences of SSC malfunctions, possibility for the creation of an accident of a different type, possibilities for malfunctions of SSC's, impact on the fuel cladding, reactor coolant pressure boundary, or containment, or changes in the method of evaluations due to the changes associated with this package. Were the 10 PSI increase not to occur and these changes were implemented the result would be a reduction in operating margin rather than safety margin. LICENSING DOCUMENT CHANGE REOUESTS (LDCR) LDCR 2003-076, Rev. 0 This proposed change is to revise U1 and U2 TS Bases to remove the discussion of the automatic swap of PSW cooling water to the "B" Control Room AC unit; add a description to the U 1 FSAR for the manual action, in place of the automatic action, for providing cooling water to the MCR AC unit 1241-BOO8B from the PSW Division 11; and add component "condensing unit," its malfunction and Enclosure Page 10 of 14 Edwin I. Hatch Nuclear Plant Report of Facilitv Changes, Tests, and Experiments Safety Evaluation Summaries comments to the MCR HVAC Systems Failure Analysis Table as the result of the manual action. This a change to the plant as described in the U1 FSAR section 10.7.6 and U1 TS Bases B 3.7.5. The automatic transfer function is merely a design feature of the B MCR AC train that is included for additional defense-in- depth. Retention of this feature is not necessary for operability of the MCR AC units. Probability of failure of the PSW function is unaffected.
The overall system operation will remain as described in the FSAR. Therefore, the probability of system failure remains unaffected as well. LDCR 2004-006, Ver. 1 I. The LDCR addresses Unit 1 TRM TLCO 3.3.1 0 in that a one-time extension of the completion time for the LC0 is being proposed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days. 2. On January 30,2004, the Unit 1 turbine "A" Master Trip Solenoid failed to function during the performance of TSR 3.3.10.1. This failure placed the unit in a required action to isolate the turbine from the steam supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The initial apparent cause of the failure is due to sticking of the solenoid.
The safety basis for the proposed one-time change is provided as follows: With the master trip solenoid inoperable, electrical overspeed protection for the turbine is not available.
However, the mechanical overspeed trip is unaffected by this component failure.
The surveillance on the mechanical overspeed trip is up-to-date.
The FSAR (HNP- 1 -FSAR-7.11.3) identifies the mechanical overspeed as the protection feature credited with preventing catastrophic overspeed of the turbine. The backup overspeed trip is credited only when the mechanical overspeed trip is locked out for testing.
Significant margin (-50%) exists between the mechanical overspeed trip setpoint and the speed at which the overstress could possibly lead to failure. The incremental probability of occurrence of an overspeed event during the seven days allowed by this change is judged to be very small.
The following compensatory measures are recommended for the seven-day period: In order to decrease even Mer the potential for grid-induced events that might result in a generator load rejection, switchyard work involving breakers should be curtailed.
No work that would result in a change in EOOS color is permitted.
Enclosure Page 1 1 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes, Tests, and Experiments Safety Evaluation Summaries Operations personnel should brief at the beginning of each shift that the Master Trip Solenoid Valve (MTSV) is unavailable and prompt operator action may be required to cause turbine trip. GENCOM should be apprised of the turbine status (without electrical overspeed protection) and asked to minimize any action that could lead to a load reject.
LDCR 2004-01 1, Rev.
1 1. Revision 1 addresses Unit 1 TRM TLCO 3.3.10 in that a one-time extension of the completion time for the LC0 is being proposed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 8 days. 2. On February 06,2004, the Unit 1 turbine "B Master Trip Solenoid failed to hction during the performance of TSR 3.3.10.1. This failure placed the unit in a required action to isolate the turbine from the steam supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The initial apparent cause of the failure is due to sticking of the solenoid.
- 3. The safety basis for the proposed one-time change is provided as follows: With the master trip solenoid inoperable, electrical overspeed protection for the turbine is not available. However, the mechanical overspeed trip is unaffected by this component failure.
The surveillance on the mechanical overspeed tip is up-to-date.
The FS AR (HNP- 1 -FS AR-7.1 1.3) identifies the mechanical overspeed as the protection feature credited with preventing catastrophic overspeed of the turbine.
The backup overspeed trip is credited only when the mechanical overspeed tip is locked out for testing. Significant margin (approximately 50%) exists between the mechanical overspeed trip setpoint and the speed at which the overstress could possibly lead to failure. The incremental probability of occurrence of an overspeed event during the 8 days allowed by this change is judged to be very small.
The following compensatory measures are recommended for the 8-day period: In order to decrease even further the potential for grid-induced events that might result in a generator load rejection, switchyard work involving breakers should be curtailed. That is, no work that would result in a change in EOOS color is permitted. Operations personnel should brief at the beginning of each shift that the MTSV is unavailable and prompt operator action may be required to cause turbine trip.
Enclosure Page 12 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes. Tests, and Exveriments Safetv Evaluation Summaries GENCOM should be apprised of the turbine status (without electrical overspeed protection) and asked to minimize any action that could lead to a load reject.
LDCR 2004-014, Rev. 0 This proposed change is to revise the Unit 1 TRM TSR 3.9.3.1 to remove the requirement to perform a hoist limit loaded interlock test for the refueling platform fuel grapple and the auxiliary hoist every 7 days after its initial performance.
This is a change to the Unit 1 TRM that relaxes the surveillance frequency requirements on the loaded interlock surveillances for the refuel platform fuel grapple and the auxiliary hoist.
The loaded interlocks setpoint surveillance in the TRM insures that the fuel grapple and the auxiliary hoist are capable of detecting when a fuel bundle has been lifted by the refuel platform. That fuel loaded signal is a part of the refueling interlocks. For example, when the fuel grapple is loaded, a rod withdrawal block will engage if the refuel platform is near the core and the mode switch is in the refuel position.
This particular TRM surveillance verifies that the setpoint on the fuel grapple loaded signal is adequate. A separate Technical Specifications surveillance (not affected by this TRM revision) will insure that the integrated signals together provide the necessary signal (in the example above, the rod block). Furthermore, the nature of the refueling interlocks is such that any problems with the interlocks will be evident to the refuel platform operator.
As a result, increased frequencies for their surveillances are of questionable value.
For example, when the fuel grapple is loaded with a fuel bundle, a "fuel grapple loaded" annunciator is provided to the operator in the platform cabin. It is therefore likely that any failure of this interlock will be obvious to the operator.
This point is noted in the Technical Specifications Bases for the refueling interlock surveillance SR 3.9.1.1 : "The 7 day frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel." No other plant systems are involved in the TRM change. For the above reasons, the likelihood of occurrence of a previously evaluated system or component malfunction is not increased.
Enclosure Page 13 of 14 Edwin I. Hatch Nuclear Plant Report of Facility Changes. Tests, and Experiments Safety Evaluation Summaries LDCR 2004-022, Ver. 2 This is a proposed change to the Unit 1 TRM and FSAR to increase the acceptance criterion on the RCIC AC inboard valve from 20 seconds to 25 seconds. The safety function of this AC valve, in conjunction with the outboard DC valve, 1 E5 1 -F008, is to automatically isolate on a RCIC steam supply line break in the reactor building. During the 2004 Spring Refueling outage, a gearing change was made to the valve operator and, as a result, the as left close stroke time increased to 19.9 seconds. This meets the acceptance criterion of 20 seconds, but obviously leaves very little margin. It is therefore desired to explore the possibility of increasing the stroke time acceptance criterion. Rupture of the RCIC steam line is one of the high energy line breaks (HELB) analyzed in the FSAR. Other examples include rupture of a main steam line, the High Pressure Coolant Injection (HPCI) steam supply line the RWCU supply line. These breaks are non-limiting with respect to fuel limits and vessel inventory because, unlike the breaks in the primary containment, they isolate.
The HELB safety analysis calculates the mass and energy escaping from the break into the reactor building from event initiation until the valve is fully closed. This mass and energy release is used to calculate the peak temperatures and pressure differentials within the reactor building and is also the basis of the environmental qualification (EQ) temperature profiles in the reactor building. The calculation for the RCIC HELB has been reviewed and it was determined that an increase in the isolation stroke time acceptance criterion from 20 to 25 seconds can be done without affecting the reactor building temperatures, pressures, or the EQ temperature profile in the reactor building.
CAUTION TAGS 1 -CA 1 E 1 1 -00 1 44. Rev. 0 1 -CA 1 E 1 1 -00 147, Rev.
0 This activity is for binding closed or "gagging" 1 El 1 -F200B and 1 El 1 -F200C, which are the minimum flow valves for RHRSW pump 1 El 1 -COO 1 B and 1E 1 1 - COO1 C. The purpose of maintaining these valves in the closed position is to ensure that sufficient flow can be developed from the RHRSW at the worst-case conditions of river level. The FSAR states that the design hction of these valves is a low-flow bypass. Since the valves will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR.
Enclosure Page 14 of 14 Edwin I. Hatch Nuclear Plant Rmort of Facility Changes. Tests, and Experiments Safety Evaluation Summaries The valves addressed by this activity cannot form any part of a sequence of events which could cause an accident.
Hence, no change to these valves could affect the probability of occurrence of an accident.
2-CA-04-2E 1 1-00059, Rev. 0 This activity is for binding closed or "gagging" 2E 1 1 -F200B and 2E11 -F200C, which are the minimum flow valves for RHRS W pump 2E 1 1 -COO 1 B and 2E 1 1 - COO1 C. The purpose of maintaining these valves in the closed position is to ensure that sufficient flow can be developed from the RHRSW at the worst-case conditions of river level. The FSAR states that the design hction of these valves is a low-flow bypass. Since the valves will be bound closed, no low-flow bypass will be available to the affected pump. This constitutes a change to the FSAR. The valves addressed by this activity cannot form any part of a sequence of events which could cause an accident.
Hence, no change to these valves could affect the probability of occurrence of an accident.