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Category:Letter type:L
MONTHYEARL-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 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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: 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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) L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For...2019-08-0505 August 2019 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For... 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@ Xcel Energym MAR 3 O 2010 L-PI-10-029 10 CFR 50 Appendix H U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Request for Revision to Reactor Vessel Material Surveillance Capsule Withdrawal Schedule for Prairie Island Nuclear Generating Plant (PINGP)
References:
- 1. Letter from Nuclear Management Company, LLC (NMC) to NRC, "Application for Renewed Operating Licenses", dated April 11, 2008, ADAMS Accession Number ML081130666.
Pursuant to 10 CFR 50 Appendix HI Section lll.B.3, Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, hereby requests approval for a revision to the PINGP, Units 1 and 2, reactor vessel material surveillance capsule withdrawal schedule. In reference 1, NMC* submitted an application for a renewed operating license which would allow PINGP operation to be extended for 20 years. To support the proposed extended period of plant operations, NSPM proposes to remove Capsule N from Unit 1 during the Spring 201 1 refueling outage and remove Capsule S from Unit 2 during the Spring 2012 refueling outage. The existing removal schedule states that the remaining capsules are in "standby". The details of the proposed schedule revision are contained in Enclosure 1 to this letter. Based on the technical evaluation in Enclosure 1, NSPM concludes that the proposed withdrawal schedule complies with the requirements of the American Society for Testing and Materials (ASTM) standard ASTM E 185-82.
To support implementation of the revised withdrawal schedule, NSPM requests approval of this proposed change by April 1, 201 1.
- On September 22, 2008, NMC transferred its operating authority to NSPM, doing business as Xcel Energy. By letter dated September 3, 2008, NSPM assumed responsibility for actions and commitments previously submitted by NMC.
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Document Control Desk Page 2 If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-388-1 121.
Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.
Mark A. Schimmel Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (1) cc: Administrator, Region Ill, USNRC Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC
Enclosure 1 Request for Revision to Reactor Vessel Material Surveillance Capsule Withdrawal Schedule for Prairie Island Nuclear Generating Plant (PINGP)
- 1. BACKGROUND Appendix H of 10 CFR Part 50 describes reactor vessel material surveillance program requirements. Paragraph (lll)(B)(3) of this Appendix states that proposed withdrawal schedules must be submitted with a technical justification per 10 CFR 50.4, and approved prior to implementation.
The design of the reactor vessel surveillance program and the withdrawal schedule must meet the requirements of the edition of American Society for Testing and Materials (ASTM) standard ASTM E 185 that is current on the issue date of the American Society of Mechanical Engineers (ASME) Code to which the reactor vessel was purchased. The applicable ASTM E 185 version for PINGP, Units 1 and 2, is ASTM E 185-70. In addition, to the extent practicable, the test procedures and reporting requirements for each capsule withdrawal must comply with ASTM E 185-
- 82. Tables 1 and 2, which were derived from the Pressure and Temperature Limits Report, show the withdrawal schedule for PINGP, Units 1 and 2, respectively, reactor vessel surveillance capsules.
Table 1 Unit 1 Reactor Vessel Surveillance Capsule Removal Schedule I
1 I
Capsule Vessel Location (degree)
~ e a d [ ~ ] Withdrawal Factor EFPY[~]
Fluenceral (nlcm2, E>1.0 MeV)
T 67 1.89 Standby -----
237 1.77 Standby -----
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RV Capsule Withdrawal Schedule Table 2 Unit 2 Reactor Vessel Surveillance Capsule Removal Schedule 1
I Vessel Location ~ e a d [ ~ ] Withdrawal ~luence[~]
Capsule (degree) Factor EFPY[~] (nlcm2, EM.0 MeV)
' N 237 1.72 Standby -----
I 57 1.72 Standby -----
L .
- a. Updated in Capsule S dosimetry analysis
- b. Effective full power years (EFPY) from plant startup
- c. Plant specific evaluation d Updated in Capsule P dosimetry analysis PROPOSED REVISION TO THE WITHDRAWAL SCHEDULE The proposed schedule is to withdraw Unit 1 Capsule N during the Spring 201 1 outage, and Unit 2 Capsule S during the Spring 2012 outage. Unit 1 Capsule T and Unit 2 Capsule N will remain in standby. As shown below in Section 3, the proposed withdrawal schedules satisfy the requirements of ASTM E 185-70, the version that was current at the time the reactor vessels were purchased and ASTM E 185-82.
Therefore, the withdrawal schedules satisfy the requirements of Appendix H to 10 CFR Part 50.
TECHNICAL JUSTIFICATION Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, submitted an application for renewal of the PINGP operating licenses (Reference I ) , which proposed to extend plant operation for an additional 20 years. To account for license renewal, NSPM proposes to withdraw one of the two remaining capsules from each unit after the capsule receives a neutron fluence equivalent to the 60-year vessel fluence. The remaining spare capsules (one per unit) will stay in the reactor vessel to provide meaningful metallurgical data for potential future license renewals.
The proposed surveillance capsule withdrawal schedule is based on the requirements specified in ASTM E 185-82, Section 7.6. Six surveillance capsules were installed in each unit at PINGP. Four capsules have been removed per the Page 2 of 4
RV Capsule Withdrawal Schedule established withdrawal schedule. The last capsule to be removed from Unit 1 was Capsule S at 18.12 EFPY and the last capsule removed from Unit 2 was Capsule P at 17.24 EFPY. In accordance with the requirements of ASTM E 185-82, Section 7.6.2, NSPM proposes to withdraw one of the remaining capsules from each unit when its neutron fluence exposure exceeds the new peak end-of-life (EOL) (54 EFPY) vessel fluence, but prior to exceeding twice that fluence exposure.
The maximum reactor vessel fluence at 54 EFPY for Unit 1 is projected to be 5.162E19 n/cm2 (Esl .O MeV), and for Unit 2 is projected to be 5.196E19 n/cm2 (E>1.0 MeV). Per the requirements of ASTM E 185-82, Section 7.6, the surveillance capsules should be removed when their neutron fluence exceeds the new peak EOL vessel fluence (i.e., 5.162E19 nlcm2for Unit 1 at 54 EFPY, and 5.196E19 nlcm2for Unit 2 at 54 EFPY), but prior to exceeding twice that fluence exposure (i.e.,
1.032E20 n/cm2for Unit I , and 1.039E20 nlcm2for Unit 2). Calculations have been performed to determine the earliest withdrawal times for the remaining capsules which are shown in Table 3.
Table 3 Earliest Withdrawal Times Removal Time Unit Capsule Lead Factor (2EFPY) 1 T 1.89 24.2 N 1.77 26.5 As of August I,2009, the total lifetime performance of Unit 1 was 30 EFPY and of Unit 2 was 29.9 EFPY. Additional calculations have been performed to estimate the neutron fluence of each capsule at the time of the proposed next withdrawal. The results are shown in Table 4.
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RV Capsule Withdrawal Schedule Table 4 Projected Fluences Projected Unit Capsule Proposed EFPY Fluence Removal (nlcm2) 1 T Standby 31.6 6.292E19 N Spring 201 1 31.6 5.893E19 2 N Standby 32.2 5.739E19 1 S I Spring 2012 1 32.2 1 5.739E19 1 After capsule removal, the surveillance specimens will be tested in accordance with the requirements of 10 CFR 50, Appendix H and ASTM E 185-82. The results of materials testing, fluence analysis, and effective full power years (EFPY) from startup are used to predict the effects of neutron embrittlement through the end of extended life.
- 4. REFERENCES
- 1. Letter from Nuclear Management Company, LLC (NMC) to NRC, "Application for Renewed Operating Licenses", dated April 11, 2008, ADAMS Accession Number ML081130666.
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