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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 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) 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... L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 L-PI-19-014, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-04-29029 April 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-PI-19-003, Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP)2019-02-0404 February 2019 Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-006, Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements2019-01-29029 January 2019 Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements L-PI-19-005, Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.692019-01-15015 January 2019 Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.69 L-PI-18-063, Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 8052018-12-0606 December 2018 Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 805 2024-01-02
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N Commrrtsd to Nucl Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-06-053 10 CFR 50.46 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie lsland Nuclear Generating Plant Unit 1 Docket 50-282 License No. DPR-42 Corrections to Emerqencv Core Cooling Svstem (ECCS) Evaluation Models
Reference:
- 1) Letter L-PI-06-010 from Nuclear Management Company, LLC to Nuclear Regulatory Commission, "Revised Commitment to Submit Best-Estimate Loss of Coolant Accident (LOCA) Analysis," dated February 2, 2006.
Enclosed is a report of changes to the Prairie lsland Nuclear Generating Plant (PINGP)
Unit 1 Emergency Core Cooling System (ECCS) Evaluation Models. This report is being submitted in accordance with the provisions of 10 CFR 50, Section 50.46, as a 30-day report.
The report includes Large Break Loss of Coolant Accident (LBLOCA) changes reported by Westinghouse. There were no changes to the Small Break Loss of Coolant Accident (SBLOCA) analysis.
The Unit 1 LBLOCA Peak Clad Temperature (PCT) changes (Enclosure 1) are due to the following two changes:
Unit 1 reactor vessel head replacement included a head assembly upgrade package (HAUP) designated as "HAUP LOCA Evaluation" (+3 degree penalty)
Reconstitution of one fuel assembly with two natural uranium rods (+I degree penalty)
The PCT (2043 OF, see Enclosure) for LBLOCA analysis continues to remain below the 10 CFR 50.46 PCT acceptance criterion. The accumulated absolute value of the PCT changes and errors since the original 1995 baseline Analysis of Record is 799 OF. NMC has committed by letter (see Reference I ) to provide a new LBLOCA analysis for PINGP by July 31, 2006. The limiting LOCA analysis for Prairie lsland Unit I , with consideration of all 10 CFR 50.46 assessments, remains the LBLOCA analysis.
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Document Control Desk Page 2 The enclosure to this letter need not be withheld from public disclosure.
Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments Thomas J. Palmisano Site Vice President, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC
ENCLOSURE NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO 50-282 Unit 1 LBLOCA Peak Clad Temperature Summaries (includes plant specific changes and non-zero non-plant specific changes) 3 pages follow
Westinghouse Proprietary Class 2 Our ref: LTR-LIS-06-277 Page 7 of 1 1 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie Island Unit 1 Utility Name: N u c l e a r M a n a g e m e n t C o m p a n y , LLC Revision Date: 5 / 12/06 Analysis Information EM: SECY UP1 Analysis Date: 3/1/95 Limiting Break S i e : C d = 0.4 FQ: 2.4 FdH: 1.77 Fuel: OFA SGTP (%): 15 Notes: Zirlom, O S G SGTP Evaluated up to 24.64% (see also Note e); Fq increased t o 2.5 (Item A.lO); RSG Study at 10% SGTP.
Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 2180 12 (a)
PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS 1 . Fixed Heat Transfer Node Ass~gnmentErrorlAccumulator Water -175 3 Injection Error (1995 Report) 2 . I-D Transition Boiling Heat Transfer Error (1 997 Report) 3 . Vessel Channel DX Enor (1997 Report) 4 . Input Consistency (1997 Report) 5 . No Items for 1996 & 1998 Reports 0 4.6 6 . Accumulator LindPressurizer Surge Line Data / Plant Specific Accumulator Level & Line Volume I Plant Specific Restart Errw.
Reanalysis ( I 999 Report) 7 . Modeling Updates and Unheated Conductor Input Corrections (Plant Specific, 2000 Report) 8 . Accumulator Pressure +I- 30 psi Range (Plant Specific, 2001 Repoit) 9 . LHSl Error Enhlation (Plant Specific, 2002 Report) 10 . Sensitlvrty Study for F e 2 . 5 , LHSl Corration, dc. (as listed in note
( f ) ) (Plant Specific, 2003 Report) 1I . Broken Loop Nozzle Loss Cwfficient (Plant Specific) -19 18,19,21, (h) 25 12 . SECY Cold Leg Nozzle Expansion 13 25 B. PLANNED PLANT MODlFlCATlON EVALUATIONS I . Sensitivity S ~ d for y Steam Generator Tube Plugging Increase to 25%
2 . Accumulator Wata Volumc +I- 25 f13 Range 3 . Accumulator Pressure Extended to +I- 55 psi Range 4 . 2 Rcconstitutcd Rods Evaluation 5 . SATP Core Average Burnup 6 . Sensitrviry S ~ d for y Framatwne Replacement Steam Generators 7 . HAUP UH3A Evaluation C. 2006 ECCS MODEL ASSESSMENTS
Westinghouse Proprietary Class 2 Our ref: LTR-LIS-06-277 Page 8 of 1 I Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie lsland Unit I Utility Name: Nuclear Management Company, LLC Revision Date: 5 /I 2/06 1 .None 0 D. OTHER*
1 . Removal of Reference 14 L.HSI Error Evaluaiion -30 16 (s)
LICENSING BASIS PCT + PCT ASSESSMENTS PCT- 2043
- It is recommended that the licensee detnmine if these K f allocations be considered with respect to 10 CFR 50.46 reporting requirements.
References:
1 .95NSG-0021, "Updated UP1 LBLOCA," March 24,1995.
2 . WCAP-13919, Addendum 1, "Prairie lsland Units 1 and 2 W C O B M R A C Best Estimate UP1 Large Break LOCA Analysis Engineering Report Addendum 1: Updated Results," December 19%.
3 . NSP-96-202, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting," February 20. 1996.
4 . NSP-97-20], "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reponing," April 17. 1997.
5 . NSP-98-012, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reponing for 1997." February 27, 1998.
6 . NSP-99-010, "Northern States Power Company Prairie lsland Unlts 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 1998," April 29,1999 7 . NSP-00-005, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reponing f a 1999," February 2000.
8 . NSP-00-057, "Northern Stales Power Company Prairie lsland Units I and 2 LOCA Evaluation of 25% SGTP with Other Modeling Updates," December l I. 2000.
9 . LTR-LIS-06-277, "Reconstitution Evaluation, 10 CFR 50.46 Reporting Plant Specific Text, and Updated Rackup S h m s for F'rairic lsland Unit 1, Cycle 24," 5i2006.
10 . NSP-01-006. "Northern States Power Company Rairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Repolting for 2000," March 6.2001.
I I . NSP-02-9, "Nuclear Management Company Prairie lsland Units 1 and 2 LBUW3A Accumulator Pressure and Volume Ranges Evaluation," February 15, 2002.
12 . NSP-02-5. "Nuclear Management Company t h i n e lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2001," March 2002.
13 . NSP-02-59lLTR-ESI-02-194, "Final Evaluation of Large Brcak LOCA Enor," December 2002.
14 . NSP-03-19, 'Wuclear Management Company Prairie bland Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.
IS . MP92-TAk-0394 1 ET-NSL-OPL-1-92-518, "NSPC Prairie lsland Units 1 and 2, SG Tube Row Area Rcduaion under LOCA I SSE - Fiml Repon", October 21,1992.
16 . NSP-04-10 "Safely Analysis Transtion Program Transmittal of Engineering Report," February 20,2004.
17 . NSP-93-513, Rev IET-NSL-OPL-1-93-313, RN. I , Lcner from T. A. Hawley (W) to K. E. Higar (NSP). "Final Transmittal of Assumptions to be used for the Large and Small Break LOCA Analyses, Rev. I ",July 7,1993. Confirmed by : Lmer from K. E. Higar (NSP) to Mr. T. Hawley (W), "Acceptance ofNSP 513, Rev. I", July 30,1993.
18 . NSP-04-38, "Nuclear Management Company Rairic lsland Units 1 and 2 10 CFK 50.46 Annual Notification and Reponing for 2003," March 2004.
19 . WCAP-162M-P, "SAW Engineering Report for Prairie Island." February 2004.
Westinghouse Proprietary Class 2 Our ref: LTR-LIS-06-277 Page 9 of 1 l 20 . NF-NMC-04-49, "Nuclear Management Company Prairie lsland Unit 1 Cycle 22 Final RSE," April 2004.
21 . NSP-04-65, "Nuclcar Management Company Prairie lsland Unlts I & 2 Safety Analyas Transillon Program Rcpsonsc lo 10 CFR 50.46 inquiry," Apnl21,2004.
22 . NF-NMC-04-129. "Nuclear Management Company Prairie lsland Unit 1, Cycle 23 Final RSE," August 2004.
23 . NSP-04-114, "Nuclear Management Company Prairie lsland Units 1 & 2, Safery Analysis Transition Program. Transmittal of LBLOCA Replacement Steam Generator (RSG)Engineering Report Addendum," (WCAP-16206-P-Addendum I), June 24 . NSP-05-155, "Nuclear Management Company, Reactor Vesxl Head Replacement Project, Prairie Island Units 1 & 2," May
- 18. 2005.
25 . NSP-05-191, "Miscellanmus LBLOCA SECY EM Enor Notification," August 2005.
Notes:
(a) P-bar-HA increaxd from 1.57 to 1.59 fb) Reanalysis for all listed i s s u s (c) Keanalys~sfor both i s s u a (d) Related ICO in existence (NSP-01-030). NMC cognizant of uncertainty application and PCT sheet categonzition.
(e) It is assumed that NMC is applying the 0.36% SGTF' allowance factor branch of the SG LOCA / SSE issue (Refermcc IS).
Thus the 25% SGTP Smdy (Item B.1) suppons a net SGTP limit of 24.64%.
(f) Sensitivity Study for: FQ2.50, PAD 4.0 Implementation, Restoration ofLHS1 to Reference 17 values, SGlLoop AP Re-hming, Core Pown Ratoration.
(g) Thc note (f) sensitivity study allows for the removal ofthe Reference 13 engineering assessment.
(h) Items A.10 and A.11 presented as aggregate -66 O F entry prior to Reference 21 decomposition.