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Category:Letter type:L
MONTHYEARL-PI-24-036, – Preparation and Scheduling of Operator Licensing Examinations2024-06-28028 June 2024 – Preparation and Scheduling of Operator Licensing Examinations L-PI-24-031, Independent Spent Fuel Storage Installation, Supplement to License Amendment Request to Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2)2024-06-0505 June 2024 Independent Spent Fuel Storage Installation, Supplement to License Amendment Request to Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2) L-PI-24-014, License Amendment Request to Revise the Technical Specification Definition of Reactor Trip System (RTS) Response Time and Apply Response Time Testing to RTS Trip Functions with Time Delay Assumption2024-06-0303 June 2024 License Amendment Request to Revise the Technical Specification Definition of Reactor Trip System (RTS) Response Time and Apply Response Time Testing to RTS Trip Functions with Time Delay Assumption L-PI-24-030, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.12024-05-22022 May 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.1 L-PI-24-012, Independent Spent Fuel Storage Installation - License Amendment Request: Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2)2024-04-0202 April 2024 Independent Spent Fuel Storage Installation - License Amendment Request: Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2) L-PI-24-004, Independent Spent Fuel Storage Installation - Annual Effluent Report, January Through December 20232024-02-29029 February 2024 Independent Spent Fuel Storage Installation - Annual Effluent Report, January Through December 2023 L-PI-24-009, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Unit 2, Alternative RR-09 Safety Injection System and Volume Control System Category C Check Valve Quarterly Testing2024-02-13013 February 2024 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Unit 2, Alternative RR-09 Safety Injection System and Volume Control System Category C Check Valve Quarterly Testing L-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report L-PI-23-010, Annual Report of Individual Monitoring2023-04-27027 April 2023 Annual Report of Individual Monitoring L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2023-03-28028 March 2023 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-005, CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2023-03-0303 March 2023 CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) L-PI-23-001, Day Steam Generator Tube Inspection Report2023-01-30030 January 2023 Day Steam Generator Tube Inspection Report L-PI-22-047, Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report2022-12-21021 December 2022 Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2022-12-0202 December 2022 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-22-040, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-10-0606 October 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts2022-09-20020 September 2022 Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts L-PI-22-032, CFR 50.46 LOCA Annual Report2022-06-16016 June 2022 CFR 50.46 LOCA Annual Report L-PI-22-033, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles2022-06-10010 June 2022 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles L-PI-22-003, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0707 June 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-024, Supplement to Application for License Amendment to Implement 24-Month Operating Cycle2022-03-0707 March 2022 Supplement to Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-047, Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 22021-12-0707 December 2021 Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 2 L-PI-21-045, Response to Request for Additional Information Cooling Water System License Amendment Request2021-11-0404 November 2021 Response to Request for Additional Information Cooling Water System License Amendment Request L-PI-21-029, Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.12021-10-0707 October 2021 Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.1 L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions2021-10-0202 October 2021 License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions L-PI-21-032, Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island2021-09-30030 September 2021 Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island L-PI-21-016, Application for License Amendment to Implement 24-Month Operating Cycle2021-08-0606 August 2021 Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-027, 2020 10 CFR 50.46 LOCA Annual Report2021-06-28028 June 2021 2020 10 CFR 50.46 LOCA Annual Report L-PI-21-023, Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report2021-05-14014 May 2021 Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report L-PI-21-007, Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes2021-04-19019 April 2021 Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes L-PI-20-050, Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic2020-10-0707 October 2020 Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic L-PI-20-051, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2020-09-28028 September 2020 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-20-026, Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiativ2020-09-0101 September 2020 Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 L-PI-20-035, = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule2020-07-28028 July 2020 = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule L-PI-20-023, Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI)2020-06-10010 June 2020 Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI) L-PI-20-014, Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI2020-04-29029 April 2020 Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI L-PI-20-004, License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.132020-03-30030 March 2020 License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.13 L-PI-20-001, License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-12020-01-29029 January 2020 License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-1 L-PI-19-041, Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2019-12-23023 December 2019 Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-19-031, License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2019-12-16016 December 2019 License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b L-PI-19-040, License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency2019-10-0707 October 2019 License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency L-PI-19-038, Submittal of Revised Pressure and Temperature Limits Report2019-09-19019 September 2019 Submittal of Revised Pressure and Temperature Limits Report L-PI-19-037, Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals2019-09-16016 September 2019 Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals L-PI-19-025, Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP)2019-08-27027 August 2019 Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP) 2024-06-05
[Table view] Category:Safety Evaluation
MONTHYEARML24071A1162024-05-0101 May 2024 Issuance of Amendment Nos. 244 and 232 Revise TS 3.7.8, Cooling Water (Cl) System ML24100A8042024-04-24024 April 2024 – Alternative Request RR-09 for Safety Injection and Volume Control System Category C Check Valve Inservice Testing ML24060A1232024-03-27027 March 2024 to Request 1-RR-5-10 and 2-RR-5-10 Regarding Reactor Pressure Vessel Welds and Nozzle Welds ML23356A0032024-01-17017 January 2024 Issuance of Amendments Revise Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) ML23215A1672023-12-15015 December 2023 Acceptance of Requested Licensing Action Amendment Request to Revise Surveillance Requirement 3.8.1.2 Note 3 ML23115A4072023-04-26026 April 2023 Correction of License Amendment Nos. 226 and 214 Adoption of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B ML22357A1002023-03-31031 March 2023 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments Standard Emergency Plan and Consolidated Emergency Operations Facility ML22300A2232022-11-0101 November 2022 Issuance of Amendments 241 and 229 TSTF-577 Revised Frequencies for Steam Generator Tube Inspections ML22270A3252022-09-30030 September 2022 (Prairie Island), Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-08 ML22181A0002022-08-17017 August 2022 Issuance of Amendments Reactor Trip System Power Range Instrumentation Channels ML22166A3892022-07-28028 July 2022 Issuance of Amendments 239 and 227 24-Month Operating Cycle ML22061A2062022-04-0101 April 2022 Issuance of Amendments TSTF-471, Rev. 1, TSTF-571-T, and Administrative Changes to Technical Specification Section 5.0 ML21312A0212021-11-23023 November 2021 Issuance of Amendment Nos. 237 and 225 Inoperable Cooling Water System Supply Header ML21008A0012021-03-19019 March 2021 Issuance of Amendment Nos. 236 and 224 Low Temperature Overpressure Protection ML20346A0202021-03-15015 March 2021 Issuance of Amendment Nos. 235 and 223, Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML20356A0022021-02-0505 February 2021 Issuance of Amendment Nos. 234 and 222, Revise Technical Specifications 3.2.1 and 5.6.5 Identified in Westinghouse Nuclear Safety Advisory Letters NSAL-09-05, Revision 1, and NSAL-15-1 ML20283A3422020-11-18018 November 2020 Issuance of Amendment Nos. 233 and 221 Adoption of Technical Specifications Task Force Traveler TSTF-547, Clarification of Rod Position Requirements ML20276A0032020-10-0202 October 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20217L1852020-10-0202 October 2020 Issuance of Amendment Nos. 232 and 220 Increase the Integrated Leak Rate Test Program Type a and Type C Test Frequency ML20210M0142020-09-0808 September 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 204, 231, and 219 TSTF-529 Clarify Use and Application Rules ML20230A0512020-09-0303 September 2020 Reactor Vessel Material Surveillance Capsule Withdrawal Schedules ML20153A4012020-06-0101 June 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20134H9582020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML19276F6842019-11-12012 November 2019 Issuance of Amendments 230 and 218 Issuance of Amendments Adoption of 10 CFR 50.69 - Risk Informed Caterization and Treatment of Structure, Systems and Components of Nuclear Power Reactors ML19232A1512019-11-0707 November 2019 Issuance of Amendments Modifying the Design Basis for Quality Classification of Certain Fuel Handling Equipment ML19140A4472019-07-30030 July 2019 Issuance of Amendments Revision to National Fire Protection Association (NFPA) Standard NFPA 805 Modifications ML19177A3802019-07-0303 July 2019 Reactor Vessel Material Surveillance Capsule Withdrawal Schedules ML19128A1332019-06-0606 June 2019 Issuance of Amendments TSTF-439 Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO ML19045A4802019-04-16016 April 2019 Issuance of Amendment Nos. 226 and 214 Adoption of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B ML19046A1662019-02-25025 February 2019 Relief from the Requirements of the ASME Code ML19029A0942019-01-29029 January 2019 Issuance of Amendment No. 213, One-Time Technical Specification Change to Extend Completion Time for EDGs D5 and D6 (Emergency Circumstances) ML18270A0152018-10-19019 October 2018 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME OM Code ML18100A7882018-05-0101 May 2018 Issuance of Amendment Special Heavy Lifting Device Nondestructive Examination Frequency (CAC Nos. MG0072 and MG0073; EPID L-2017-LLA-0280) ML17346A3612018-03-0606 March 2018 Issuance of Amendment Nos. 224 and 211 to Adopt Changes to the Emergency Plan ML17362A2022018-03-0505 March 2018 Issuance of Amendment Concerning Revision to the Prairie Island Nuclear Generating Plant, Units 1 and 2 Emergency Plan (CAC Nos. MF9345 and MF9346; EPID L-2017-LLA-0175) ML17334A1782017-11-30030 November 2017 Issuance of Amendment Request Related to Spent Fuel Pool Criticality Technical Specification Changes (CAC Nos. MF7121 and MF7122, EPID L-2015-LLA-0002) Non-Proprietary ML17310B2392017-11-28028 November 2017 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Unit Staff Qualifications (CAC Nos. MF9545, MF9546, and MF9547; EPID L-2017-LLA-0195) ML17163A0272017-08-0808 August 2017 Issuance of Amendments Transition to NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants ML17130A7162017-06-20020 June 2017 Issuance of Amendments Technical Specification 3.8.7 Inverters-Operating ML17110A2752017-05-0404 May 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML16320A0212016-11-28028 November 2016 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Review of Changes to the Northern States Power Company Quality Assurance Topical Report ML16256A5142016-10-13013 October 2016 Issuance of Amendment One-Time Extension for Technical Specification Surveillance Requirement 3.8.4.3. DC Sources - Operating ML16246A2002016-10-0404 October 2016 Request for Relief 1RR-4-11 and 2-RR-4-11 Associated with Certain Inservice Inspection of Component Welds for the Fourth 10-Year Interval ML16133A4062016-06-16016 June 2016 Issuance of Amendment Nos. 217 and 205, Adoption of Technical Specifications Task Force TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation (CAC Nos. MF6449 and MF6450) ML15264A2092015-11-30030 November 2015 Issuance of Amendment Nos. 216 and 204 Regarding Revisions to Technical Specification 3.3.3 and Renewed Facility Operating License ML15229A1762015-08-26026 August 2015 Issuance of Amendment Nos. 215 and 203 Regarding Revision to Licensing Basis Analysis for a Waste Gas Decay Tank Rupture ML15196A2132015-08-17017 August 2015 Requests 1-RR-4-9 and 2-RR-4-9 Associated with the Fourth Ten-Year Interval for the Inservice Inspection Program ML15196A2412015-08-12012 August 2015 Requests 1-RR-5-6 and 2-RR-5-6 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program ML15196A2272015-08-12012 August 2015 Requests 1-RR-5-3 and 2-RR-5-3 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program 2024-05-01
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N Comrnrtted to Nuclear Excellent Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-07-074 10 CFR 50.59 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units Iand 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 50.59 EVALUATION
SUMMARY
REPORT With this letter, the Nuclear Management Company, LLC, (NMC) submits two enclosures. Enclosure 1 contains descriptions and summaries of safety evaluations for changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. contains discussion of changes to regulatory commitments made within our Regulatory Commitment Change Process during the period since the last update.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
Michael D. Wadley U Site Vice President, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS - DECEMBER 2007 Below are a brief description and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out at the Prairie Island Nuclear Generating Plant by Nuclear Management Company, LLC (NMC) without prior Nuclear Regulatory Commission (NRC) approval, pursuant to the requirements of 10 CFR 50.59.
50.59 Evaluation No. 1038 - Use of Ultimate Strength Design (USD) Methodology to Evaluate Vertical Seismic Loads on Floors Description of Change This non-design change pertains to the use of different methods for evaluating seismic loads than those pre-scribed in the Prairie Island Updated Safety Analysis Report (USAR). The methods of evaluation were used in the following four calculations:
- 1. Calculation S-B01-VS-001, Evaluation of Auxiliary Building Floors for Vertical Seismic Loads
- 2. Calculation S-B01-VS-002, Evaluation of Unit 1 Reactor Building Floors for Vertical Seismic Loads
- 3. Calculation S-BOA-VS-003, Evaluation of Unit 1 Reactor Building Floors for Vertical Seismic Loads
- 4. Calculation S-B01-VS-004, Evaluation of Screenhouse Floors for Vertical Seismic Loads In each of the above calculations, concrete beams and slabs were analyzed using Ultimate Strength Design (USD) methodology versus Working Stress Design (WSD) methodology prescribed in USAR Table 12.2-6, "Allowable Stresses - Reinforced Concrete," and elsewhere.
Summary of 50.59 Evaluation Because the activity described in this evaluation is four structural analyses versus a physical change, Questions #I thru #7 in the Basis of Determination part of this evaluation were determined not to be applicable. Specifically, additional accidents or malfunctions, different accidents or malfunctions, more severe accidents or malfunctions, etc., due to one or more physical changes are not possible.
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Question #8 was determined to be applicable. The response to this question concluded that, though the four structural calculations evaluated herein used methodologies different from that described in USAR Section 12, the USD methodology used in the four analyses was in accordance with Sections 3.8.3 & 3.8.4 of NUREG-0800, Standard Review Plan, and was formally endorsed by the NRC in an April 1992 Safety Evaluation Report applicable to the Prairie Island D5lD6 Building. This 50.59 Evaluation determined that it would be appropriate to use this same methodology to evaluate the structural adequacy of existing concrete floors in four other Design Class 1 structures.
50.59 Evaluation No. 1039 - Revised Auxiliary Building High Energy Line Break Analysis Description of Change The change being evaluated is a new Auxiliary Building High Energy Line Break Analysis utilizing mass and energy releases generated by Westinghouse and new initial compartment temperatures. This analysis encompasses mass and energy release from both the Framatome Replacement Steam Generators as well as the original Westinghouse Steam Generators. The analysis determined the peak compartment pressures as well as the temperature profile. The pressure results were utilized to ensure the structural integrity of the concrete block walls and steam exclusion doors.
The temperature profile was used to ensure the Environmental Qualifications for equipment needed to mitigate the High Energy Line Break was satisfactory.
Summary of 50.59 Evaluation There are no physical or operational changes being made to any plant equipment.
Thus there is no impact on the frequency of occurrence of an accident, nor for an accident of a different type. The resultant Auxiliary Building pressures following a High Energy Line Break satisfy all applicable acceptance criteria and the resultant temperatures are bounded by the equipment qualification records. Thus there is no impact on the likelihood of a malfunction, nor for a malfunction with a different result, nor an increase in consequences of an accident or malfunction. The activity does not involve a design basis limit for a fission product barrier. The methodology used in the analyses has been reviewed against methods described in the USAR and methods approved for other facilities. Based on this review it was concluded that the methodology is not a departure from that described in the Updated Safety Analysis Report.
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50.59 Evaluation No. 1050 - Revised Small Break LOCA Analysis Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSl Condensation Model (WCAP-10054-P-A Add. 2 Rev. 1)
Description of Change This 50.59 evaluation involves a change in methodology only. It proposes to officially recognize the revision to the Small Break Loss of Coolant Accident (SBLOCA) methodology described in WCAP 10054-P-A, Addendum 2 Revision 1. The Nuclear Regulatory Commission (NRC) has expressed concerns about a portion of this methodology regarding removing the loop seal clearing restriction for the intact loop. As a result, Prairie lsland will adopt all portions of this methodology with the exception that the loop seal clearing restriction will be kept for the intact loop (only the broken loop loop seal will be allowed to fully clear).
The revised methodology, as documented in WCAP 10054-P-A, Addendum 2 Revision 1, was approved by the NRC for use at other Westinghouse PWRs (Pressurized Water Reactors) and has been reviewed for applicability to Prairie Island. All the input assumptions of this revised methodology as well as conditions and limitations bound Prairie lsland Units 1 and 2. Therefore, this new methodology is acceptable for use for Prairie lsland Units 1 and 2.
Summaw of 50.59 Evaluation This proposed change is only a change to a methodology, so only question 8, "Does the activity result in a departure from a method of evaluation described in the USAR, or any pending submittal, used in establishing the design bases or in the safety analysis?" of the Basis Of Determination applies.
The proposed change modifies the SBLOCA methodology. The revised methodology has been approved for other PWRs (Kewaunee) and the NRC acknowledged that the new methodology is applicable to Westinghouse PWR designs including those with UP1 (Upper Plenum Injection). Prairie lsland Units 1 and 2 meet all the input assumptions of the new methodology as well as conditions and limitations stated in WCAP 10054-P-A, Addendum 2, Revision 2. Therefore, the change in methodology does not result in a departure from a method of evaluation as described in USAR Section 14.7 used in establishing the design bases or the safety analyses.
50.59 Evaluation No. 1054 - Unit 1 Cycle 24 Core Reload, Revisions 0 through 3 Description of Change This design change is required to allow for continued power operation of Prairie lsland Unit 1 for approximately 18 months. The fuel in the current core will be burned to a state that no longer allows for significant full power operation. This reload will replace burned fuel from Unit 1 Cycle 23 with 48 fresh fuel assemblies. This will allow the Unit 1 reactor to produce power at its rated capacity.
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Summary of 50.59 Evaluation The USAR Chapter 14 evaluations performed by NMC's Nuclear Analysis Department (NAD) and Westinghouse demonstrate that the Prairie Island Unit 1 Cycle 24 reload design and associated Core Operating Limits Report (COLR) do not result in the accepted safety limits for any accident being exceeded. The Cycle 24 design is consistent with the description of the core in the USAR. The core contains 121 fuel assemblies using a 14 x 14 fuel rod array, with 29 control rods in the same locations as described in the USAR. The only change from Cycle 23 is the distribution of new and used assemblies. This results in a redistribution of the isotopic distribution of the core that changes the core physics parameters of the reactor. The effect of these changes in the cycle physics parameters on cycle operation and accident analyses have been evaluated using NRC approved methods.
The accident analyses show that no design limits are exceeded during any analyzed transient for the cycle as designed. The cycle does not exceed any fuel burnup limits.
Therefore the reload modification for Unit 1 Cycle 24 is safe and consistent with Prairie Island's current Licensing Basis.
50.59 Evaluation No. 1055 - Unit 2 Cycle 24 Core Reload, Revisions 0 and 1 Description of Change This design change is required to allow for continued power operation of Prairie Island Unit 2 for approximately 21 months. The fuel in the current core will be burned to a state that no longer allows for significant full power operation. This reload will replace burned fuel from Unit 2 Cycle 23 with 56 fresh fuel assemblies. This will allow the Unit 2 reactor to produce power at its rated capacity. Revision 0 is valid only for Modes 5 and 6.
Summary of 50.59 Evaluation The USAR Chapter 14 evaluations performed by NAD and Westinghouse demonstrate that the Prairie Island Unit 2 Cycle 24 reload design and associated COLR do not result in the accepted safety limits for any accident being exceeded. The Cycle 24 design is consistent with the description of the core in the USAR. The core contains 121 fuel assemblies using 14 x 14 fuel rod array, with 29 control rods in the same locations as described in the USAR. The only change from Cycle 23 is the distribution of new and used assemblies. This results in a redistribution of the isotopic distribution of the core that changes the core physics parameters of the reactor. The effect of these changes in the cycle physics parameters on cycle operation and accident analyses have been evaluated using NRC approved methods. No analysis needed to be re-run for this core design.
The accident analyses show that no design limits are exceeded during any analyzed transient for the cycle as designed. The cycle does not exceed any fuel burnup limits.
Therefore the reload modification for Unit 2 Cycle 24 is safe and consistent with Prairie Island's current licensing basis.
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50.59 Evaluation No. 1056 - Mode 4 LOCA Technical Specification Bases and Procedural Changes Description of Chanae Activity
Description:
Revise Emergency Procedures 1E4 [2E4] "Core Cooling Following Loss of RHR (Residual Heat Removal) Flow" to rely on the SI subsystem as the primary source of injection during a LOCA in Mode 4 with RHR system as a backup.
Revise TS Bases B 3.5.3 to clarify that an operable train of Emergency Core Cooling System consists of the SI subsystem for injection flow and the RHR subsystem for recirculation.
Summaw of 50.59 Evaluation The changes to the emergency response procedures and TS Bases do not affect how the equipment is operated, its reliability, or performance; thus there is not impact on the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result. The evaluation demonstrates that the SI subsystem is capable of removing decay heat and thus the consequences of a LOCA in Mode 1 continues to bound those of a LOCA in Mode 4. The changes do not involve a design basis limit for a fission product barrier nor a method of evaluation. Therefore, the changes meet the design and license basis requirements.
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ATTACHMENT 2 CHANGES TO REGULATORY COMMITMENTS
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Regulatory Commitment Change 05-03 Change Frequency of Residual Heat Removal (RHR) Pump Venting in Reduced Inventory Change the RHR suction venting time from every one hour to every six hours. Original commitment was made to ensure air did not build up in the RHR pump suction.
Experience has shown that very little air builds up in an hour and that extending the venting frequency is acceptable.
Regulatory Commitment Change 06 Test Frequency Flexibility for Unit 1 Containment Fan Coil Unit Motor-Operated Valves (MOVs)
Nuclear Management Company, LLC (NMC) Committed to the Joint Owners Group (JOG) MPR-18-07 guidance for MOV testing. JOG guidance would have MOVs: MV-32133, MV-32139,MV-32142, MV-32379, and MV-32380 tested on a frequency not to exceed ten years. The ten-year frequency expires on 2/12 and 2113106 for these MOVs.
Currently, testing at power is not desirable due to the risk of pressure locking. This change would be a one time extension of the ten-year test frequency of these MOVs -
testing completed in May 2006 Unit 1 refueling outage.
Regulatory Commitment Change 06 Change Due Date of Limited Exam Relief Requests Original commitment was to submit Limited Exam Relief Requests with the outage summary report for each limited examination. The change was that Limited Exam Relief Requests will be submitted no more than 12 months after each applicable outage for each limited examination. The reason for the change is that 4th Interval Inservice Inspection is risk-based, so external analysis will be required for all components for which only a limited examination was performed and no prior examination history exists.
The 12 month due date will not go beyond the NRC requirement (which is 12 months after the interval).
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Regulatory Commitment Change 06-23 Cancel Commitment on Work Control Process Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to revise administrative procedures for the work control process to require consideration of potential effects of work on nearby equipment. This commitment was cancelled Page 1 of 3
because it was deemed unnecessary (hot work process controls have been in place nearly 30 years and will remain regardless of whether there is an associated commitment).
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Regulatory Commitment Change 06-24 Cancel Commitment on Work Control Process and Fire Protection Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to revise administrative procedures for the work management process to assure consideration is given to the need for fire prevention or suppression or equipment. This commitment was cancelled because it was deemed unnecessary (hot work process controls have been in place nearly 30 years and will remain regardless of whether there is an associated commitment).
Regulatory Commitment Change 06 Cancel Commitment on Modification Process and Fire Protection Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to revise administrative procedures for the modification process to issue a work instruction for modifications designed by offsite organizations, including responsibilities to control installation work. This commitment was cancelled because current modification process controls are better than those of 1975; thus, it is reasonable to assert that such controls will remain to meet the intent of this commitment, even if the commitment no longer exists. That is, this commitment is to do something so fundamental to the modification process in the current day that a commitment is not necessary.
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Regulatory Commitment Change 06-26 Cancel Commitment on Fire Protection Training Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to establish a method for indoctrination of offsite personnel on permanent fire protection administrative policies and procedures. This commitment was cancelled because the current badge training process is far beyond what the indoctrination training of 1975 would have been. Current training includes the necessary fire protection training (including hot work permits, combustible source use permits, fire alarms, and extinguisher types) for workers badged for unescorted access (any other "offsite" worker would have to be escorted by a badgedltrained worker). The elimination of this commitment will not result in the elimination of fire protection training.
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Regulatory Commitment Change 06-27 Cancel Commitment on Fire Protection and Work Control Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to revise the work control process to give specific consideration to Combustible Materials, Ignition Sources, Safety Monitoring Personnel, Training of Personnel, and Work Deferral. This commitment was cancelled because the training piece of this commitment is not needed (current training practices are adequate without a commitment and will remain so) and because the remainder of this commitment is addressed in later commitments.
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Regulatory Commitment Change 06 Cancel Commitment on Fire Protection and Indoctrination Training Original commitment (reconstituted from a December 30, 1975 letter to the NRC) was to establish a method for indoctrination of offsite personnel on permanent fire protection administrative policies and procedures. This commitment was cancelled to the new process - current badge training process is far beyond what the indoctrination training of 1975 would have been. Current training includes the necessary fire protection training (including hot work permits, combustible source use permits, fire alarms, and extinguisher types) for workers badged for unescorted access (any other "offsite" worker would have to be escorted by a badgedltrained worker). The elimination of this commitment will not result in the elimination of fire protection training Regulatory Commitment Change 06 Cancel Commitment on Fire Brigade Training Original commitment (reconstituted from a June 30, 1975 letter to the NRC) was to provide training for fire teams on the use of fire fighting equipment, including the use of water on electrical fires. This commitment was cancelled because fire brigade training is required, whether or not this commitment exists. In the 30 years since the original commitment was made, the means for fighting electrical fires have been well incorporated into fire brigade training.
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Regulatory Commitment Change 06-32 Change Frequency of Undervessel Bare Metal Visual (BMV) Inspection Original commitment (from response to NRC Bulletin 2003-02) was to perform a 100%
bare-metal visual exam of the lower reactor pressure vessel (RPV) dome up to and including each bottom-mounted instrumentation (BMI) penetration to RPV junction. This examination will be completed on each unit during refueling outages subsequent to the current Unit 2 refueling outage. This commitment was revised to change the frequency to every other refueling outage. The change in frequency is warranted based on site and industry experience during BMV inspections and reduction in radiation dose.
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Regulatory Commitment Change 07-01 Change Commitment on Fire Damper Installation Original commitment described in the NRC safety evaluation as, "The Licensee has committed to the following modifications: (1) A concrete fire barrier will be placed in the pipe trench that passes through the auxiliary feedwater pump rooms at the boundary between the two rooms. The existing grating will be notched and a 114-inch thick checkered floor plate will be tack welded in place to provide resistance to buckling. (2)
Fire-rated dampers (3-hour or equivalent) will be installed in all return ventilation ducts that penetrate the boundaries of the rooms." The revised commitment deleted the second item. Fire modeling demonstrated that the physical layout of the ductwork and rooms prevents damage in the auxiliary feedwater pump room from a fire originating in the 480V normal switchgear rooms.
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