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Category:Report
MONTHYEARML23270B9022023-09-29029 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23265A2532023-09-26026 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23075A3512022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 6 of 7) ML23075A3482022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 3 of 7) ML23075A3472022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 2 of 7) L-PI-23-002, WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 7 of 7)2022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 7 of 7) ML23075A3462022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 1 of 7) ML23075A3502022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 5 of 7) ML23075A3492022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 4 of 7) ML22003A1832022-01-0303 January 2022 Refueling Outage Unit 2 R32 Owner'S Activity Report for Class 1, 2, 3 and Mc Inservice Inspections ML21285A2812021-10-12012 October 2021 Revised Pressure and Temperature Limits Report ML20272A2932020-09-28028 September 2020 (PINGP) Unit 1 and 2 Revised Pressure and Temperature Limits Report ML20265A0892020-09-15015 September 2020 Draft License Conversation Record 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 ML17279A1242017-09-30030 September 2017 Enclosure 5 to L-PI-17-041, Westinghouse WCAP-17400-NP, Supplemental 1, Revision 2, Spent Fuel Pool Criticality Safety Analysis Supplemental Analysis Including the Storage of Ifba Bearing Fuel L-PI-16-058, Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation2016-07-22022 July 2016 Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation L-PI-16-054, Pressure and Temperature Limits Report, Revision 52016-06-22022 June 2016 Pressure and Temperature Limits Report, Revision 5 L-PI-16-051, 10 CFR 50.46 Emergency Core Cooling System Annual Report2016-06-22022 June 2016 10 CFR 50.46 Emergency Core Cooling System Annual Report L-PI-15-034, Pressure and Temperature Limits Report (PTLR) Revision 42015-05-14014 May 2015 Pressure and Temperature Limits Report (PTLR) Revision 4 ML15037A4582015-03-0606 March 2015 Staff Assessment of the Aging Management Program for Reactor Vessel Internal Components L-PI-14-131, Fifth Ten-Year Interval Snubbers Testing Program2014-12-18018 December 2014 Fifth Ten-Year Interval Snubbers Testing Program ML17297A3232014-11-14014 November 2014 Enclosure 2 (Redacted): Seismic Walkdown Report, in Response to the 50.54(f) Information Request Re Fukushima Near-Term Task Force Recommendation 2.3: Seismic Updated Transmittal for Prairie Island Unit 1 ML16005A1102014-09-25025 September 2014 Redacted 2014 Decommissioning Cost Analysis for the Prairie Island Nuclear Generating Plant ML14148A4772014-06-17017 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 ML14120A1622014-05-0909 May 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 L-PI-14-045, Enclosure to L-PI-14-045, Transition Report, Revision 12014-04-30030 April 2014 Enclosure to L-PI-14-045, Transition Report, Revision 1 L-PI-14-028, PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-03-27027 March 2014 PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML14030A5402014-02-27027 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14041A2042014-02-26026 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Prairie Island Nuclear Generating Plant, Units 1 and 2, TAC Nos.: MF0834 and MF0835 L-PI-13-080, First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2013-08-26026 August 2013 First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events L-PI-12-108, Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident L-PI-12-103, Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident ML12278A4052012-09-28028 September 2012 Prairie Island, Units 1 and 2, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors ML13133A0632012-06-27027 June 2012 H4, Rev. 27, Offsite Dose Calculation Manual (Odcm). ML12159A2562012-06-11011 June 2012 Review of 60-Day Response to Request for Information Regarding Recommendation 9.3 of the Near-Term Task Force Related to the Fukushima Daiichi Nuclear Power Plant Accident L-PI-10-076, Calculation No. GEN-PI-083, Revision 1, Locked Rotor Accident (LRA) Analysis Using AST, Attachment 52010-07-23023 July 2010 Calculation No. GEN-PI-083, Revision 1, Locked Rotor Accident (LRA) Analysis Using AST, Attachment 5 ML1021002592010-06-30030 June 2010 Seismic Fragilities for Unit #1 and Unit #2 Turbine Building Piping and Equipment ML1016901712010-06-11011 June 2010 Enclosure 6, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16275, Effects of Pipe Whip Interactions for Various Pipe Combinations for Internal Flooding Sdp. ML1016901702010-06-11011 June 2010 Enclosure 5, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16270, Screening of Pipe Whip Interactions for Sdp. ML1016901682010-06-11011 June 2010 Enclosure 3, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16090, Turbine Building Flooding SDP: Cl Turbine Building Pipe Break Analysis. ML1016901692010-06-10010 June 2010 Enclosure 4, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16154, Turbine Building Flooding SDP: Cl Turbine Building Seismic Pipe Break Analysis. L-PI-10-005, Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems2010-02-18018 February 2010 Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems ML1019703862009-12-31031 December 2009 Ground Water Investigation: an Improved Flow Net to Evaluate Pathways for a Potential Ground Water Release ML1002001312009-12-21021 December 2009 Report No. 0900634.401, Revision 2, Updated Leak-Before-Break Evaluation for Several RCS Piping at Prairie Island Nuclear Generating Plant, Units 1 & 2. ML1002001322009-12-18018 December 2009 Report 0900634.402, Revision 2, Updated Leak-Before-Break Report for Prairie Island Nuclear Generating Plant Unit 2 Pressurizer Surge Line Nozzle. L-PI-09-115, Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology2009-10-27027 October 2009 Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology ML1008406402009-09-11011 September 2009 Advisory Brief of Prairie Island Nuclear Generating Plant Study Group to State of Minnesota, Office of Administrative Hearings for the Public Utilities Commission, Sept. 11, 2009. Submitted with Comments on Draft Generic Environmental Impac L-PI-09-021, 2008 Unit 2 180-Day Steam Generator Tube Inspection Report2009-04-27027 April 2009 2008 Unit 2 180-Day Steam Generator Tube Inspection Report ML1020302362008-12-31031 December 2008 State of Wisconsin Prairie Island Environmental Radioactivity Survey ML0834701962008-11-21021 November 2008 PINGP - License Renewal; Radon Health Risks 2023-09-29
[Table view] Category:Miscellaneous
MONTHYEARML23270B9022023-09-29029 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23265A2532023-09-26026 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML22003A1832022-01-0303 January 2022 Refueling Outage Unit 2 R32 Owner'S Activity Report for Class 1, 2, 3 and Mc Inservice Inspections L-PI-16-058, Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation2016-07-22022 July 2016 Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation L-PI-16-051, 10 CFR 50.46 Emergency Core Cooling System Annual Report2016-06-22022 June 2016 10 CFR 50.46 Emergency Core Cooling System Annual Report ML15037A4582015-03-0606 March 2015 Staff Assessment of the Aging Management Program for Reactor Vessel Internal Components L-PI-14-131, Fifth Ten-Year Interval Snubbers Testing Program2014-12-18018 December 2014 Fifth Ten-Year Interval Snubbers Testing Program ML14148A4772014-06-17017 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 ML14120A1622014-05-0909 May 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 L-PI-14-028, PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-03-27027 March 2014 PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident L-PI-13-080, First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2013-08-26026 August 2013 First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events L-PI-12-103, Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident L-PI-12-108, Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML12278A4052012-09-28028 September 2012 Prairie Island, Units 1 and 2, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors ML12159A2562012-06-11011 June 2012 Review of 60-Day Response to Request for Information Regarding Recommendation 9.3 of the Near-Term Task Force Related to the Fukushima Daiichi Nuclear Power Plant Accident ML1016901682010-06-11011 June 2010 Enclosure 3, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16090, Turbine Building Flooding SDP: Cl Turbine Building Pipe Break Analysis. ML1016901702010-06-11011 June 2010 Enclosure 5, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16270, Screening of Pipe Whip Interactions for Sdp. ML1016901712010-06-11011 June 2010 Enclosure 6, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16275, Effects of Pipe Whip Interactions for Various Pipe Combinations for Internal Flooding Sdp. ML1016901692010-06-10010 June 2010 Enclosure 4, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16154, Turbine Building Flooding SDP: Cl Turbine Building Seismic Pipe Break Analysis. L-PI-10-005, Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems2010-02-18018 February 2010 Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems ML1019703862009-12-31031 December 2009 Ground Water Investigation: an Improved Flow Net to Evaluate Pathways for a Potential Ground Water Release ML1002001312009-12-21021 December 2009 Report No. 0900634.401, Revision 2, Updated Leak-Before-Break Evaluation for Several RCS Piping at Prairie Island Nuclear Generating Plant, Units 1 & 2. ML1002001322009-12-18018 December 2009 Report 0900634.402, Revision 2, Updated Leak-Before-Break Report for Prairie Island Nuclear Generating Plant Unit 2 Pressurizer Surge Line Nozzle. L-PI-09-115, Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology2009-10-27027 October 2009 Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology ML1008406402009-09-11011 September 2009 Advisory Brief of Prairie Island Nuclear Generating Plant Study Group to State of Minnesota, Office of Administrative Hearings for the Public Utilities Commission, Sept. 11, 2009. Submitted with Comments on Draft Generic Environmental Impac L-PI-09-021, 2008 Unit 2 180-Day Steam Generator Tube Inspection Report2009-04-27027 April 2009 2008 Unit 2 180-Day Steam Generator Tube Inspection Report ML1020302362008-12-31031 December 2008 State of Wisconsin Prairie Island Environmental Radioactivity Survey ML0834701962008-11-21021 November 2008 PINGP - License Renewal; Radon Health Risks ML0834000702008-10-24024 October 2008 PINGP License Renewal: Wisconsin Department of Health Prairie Island Environmental Survey 2006. Reference 2008c L-PI-07-057, Corrections to Emergency Core Cooling System (ECCS) Evaluation Models2007-08-23023 August 2007 Corrections to Emergency Core Cooling System (ECCS) Evaluation Models IR 05000282/20053012005-09-30030 September 2005 Er 50-282-05-301 (DRS) and 50-306-05-301 (Drs), on 08/09-19/05, Nuclear Management Company, LLC, Prairie Island Nuclear Generating Plant. Initial License Examination Report ML0520901582005-08-0202 August 2005 2004 External Stakeholder Response; 2004 Reactor Oversight Process External Survey - Attachment L-HU-05-012, Report of Unsatisfactory Blind Performance Testing Results, Monticello Nuclear Generating Plant & Prarie Island Nuclear Generating Plant, Units 1 and 22005-06-0303 June 2005 Report of Unsatisfactory Blind Performance Testing Results, Monticello Nuclear Generating Plant & Prarie Island Nuclear Generating Plant, Units 1 and 2 ML0522700732005-04-18018 April 2005 Event Notification Report for April 18, 2005 ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement ML0514604942004-04-0909 April 2004 Licensee Root Cause Report, CAP55527, Industrial Safety Issues and Poor Work Practices During Nozzle Dam Installation, with R. Alexander'S (Riii) Notes ML0318909462003-07-10010 July 2003 Relaxation of the Order, Exercising Enforcement Discretion, and Extension of the Time to Submit an Answer or Request a Hearing Regarding Order EA-03-038, Fitness-for-Duty Enhancements for Nuclear Security Force Personnel for Duane Arnold, K ML0310401092003-04-0404 April 2003 Steam Generator (#22) & Pressurizer Flaw Evaluations ML0205206392002-01-25025 January 2002 Fqa Methodology & Shutdown Margin During Physics Tests Methodology 2023-09-29
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(l Xcel Energy L-PI-16-051 JUN 2 2 2016 10 CFR 50.46 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 10 CFR 50.46 ECCS Annual Report
References:
- 1) K. Davison, PINGP, letter to NRC Document Control Desk, 201410 CFR 50.46 LOCAAnnual Report, L-PI-15-061, 6/30/2015 (ADAMS Accession No. ML15181A080)
Pursuant to 10 CFR 50.46(a)(3)(ii), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM") submits the annual report of changes to or errors in and estimated effects on the Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2 Emergency Core Cooling System (ECCS) analyses (Enclosure 1).
The limiting loss of coolant accident (LOCA) analysis for PINGP Units 1 and 2 is the large break LOCA (LBLOCA) analysis. The peak cladding temperature (PCT) for the LBLOCA analysis is unchanged for PINGP Units 1 and 2 since the last annual report (Reference 1).
Summary of Commitments This letter contains no new commitment and no revision to an existing commitment.
Scott Northard Acting Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure (1) 1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
NRC Document Control Page 2 cc: Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC
ENCLOSURE 1 10 CFR 50.46 LOCA Annual Report 8 pages follow
Westinghouse Non-Proprietary Class 3
@ Westinghouse Westinghouse Electric Company Engineering Center of Excellence 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-5598 e-mail: mcmillh@westinghouse.com Our ref: LTR-LlS-16-40 February 18,2016 Prairie Island Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2015
Dear Sir or Madam:
This is a notification of 10 CFR 50.46 reporting information pertaining to the Westinghouse Electric Company Evaluation Models/analyses. As committed to in WCAP-13451, Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting, Westinghouse is providing an Annual Report for Emergency Core Cooling System (ECCS)
Evaluation Model changes and errors for the 2015 model year. All necessary standardized reporting pages for any changes and errors for the Evaluation Models utilized for your plant(s) are enclosed, consistent with the commitment following the NUPIC audit in early 199?. Peak Clad Temperature (PCT) sheets are enclosed. All necessary revisions for any non-zero, non-discretionary PCT change to Section C have been included. Non-discretionary PCT impacts pf ooF will generally not be presented on the PCT sheet. Any plant-specific errors in the application of the model for 2015 will also be provided in Section C with discussion enclosed or cited. The Evaluation Model changes and errors (except any plant-specific errors in the application ofthe model) will be provided to the NRC via Westinghouse letter.
This information is for your use in making a determination relative to the reporting requirements of 10 CFR 50.46.
The information that is provided in this Jetter was prepared in accordance with Westinghouse's Quality Management System (QMS). Ple!!-se contact your LOCA plant cognizant engineer (PCE), Dania] Utley (412-3 74-6663), if there are any questions concerning this information.
Author: (Electronically Approved)* Verified: (Electronically Approved)*
Heather McMillen Dania! W. Utley LOCA Integrated Services II LOCA Integrated Services II Approved: (Electronically Approved)*
Matthew B. Cerrone Manager, LOCA Integrated Services II
Attachment:
10 CFR 50.46 Reporting Text and PCT Summary Sheets (7 Pages) .
- Electronically appro1*ed records are authenticated in the elecn*onic document management system.
© 2016 Westinghouse Electric Company LLC All Rights Resen*ed Page 14 of 36
.Attachment"to LTR-LIS-16-40 February 18,2016 Page 1 of7 GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses.
Examples of these changes include modifYing input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 ofWCAP-13451.
Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of 0°F.
Page 15 of36
Attachment to LTR-LIS-16-40 February 18, 2016 Page 2 of7 PRAIRIE ISLAND UNIT 1 EARLIER CONTAINMENT SPRAY ACTUATION TIMES DUE TO DEGRADED FAN COOLER HEAT REMOVAL CAPACITY
Background
Due to the design of the Prairie Island emergency core cooling system (ECCS), a complete inte1mption in high head safety injection (HHSI) flow can occur during the switchover to sump recirculation for the small break Joss-of-coolant accident (SBLOCA) transient. The Prairie Island Unit 1 SBLOCA analysis of record (AOR) models the actuation timing of the containment spray system based on containment pressurization calculations for the smaller break sizes considered therein (1.5-inch and 2-inch breaks).
This ultimately- dictates the timing of the switch over to sump recirculation and corresponding HHSI flow interruption.
Prairie Island Unit 1 identified a potential degradation in containment fan cooler heat removal capacity.
Updated containment pressurization calculations result in earlier containment spray actuation times due to the degraded heat removal capacity of the fan coolers.
This scenario is characterized as a change in plant configuration, distinguished from an evaluation model change in Section 4 of W~AP-13451.
Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect When considering the reduced heat transfer capability of the fan coolers, the containment spray pumps actuate earlier than modeled in the AOR for the 1.5-inch and 2-inch breaks; however, the inte1Tuption during the switchover to sump recirculation still occurs after the core has recovered. There is sufficient reactor coolant system inventory during the flow interruption such that a post-switchover core uncovery would not occur. Therefore, this issue is estimated as a 0°F peak cladding temperature (PCT) impact on the Prairie Island Unit 1 SBLOCA AOR.
Page 16 of36
Attachment to L TR-LIS-16-40 February 18,2016 Page 3 of?
Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best .Estimate Large Break Plant Name: Prairie Island Unit 1 Utility Name: Xcel Energy, Inc Revision Date: 2/112016 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP(%): 10 Notes:
Clad Temp (OF.) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 1 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS I . Evaluation of Fuel Pellet Thennal Conductivity Degradation and Peaking 227 2 (a)
Factor Bumdown 2 . Revised Heat Transfer Multiplier Distributions -2 3 3 . Error in Burst Strain Application 25 4 B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . None 0 C. 2015 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*
I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2015 It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.
References 1 . WCAP-17783-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Prairie Island Units 1 and 2 with Replacement Steam Generators Using ASTRUM Methodology," June 2013 ..
2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thennal Conductivity Degradation and Peaking Factor Burndown," September 20,2012.
3 . LTR-LIS-13-366, Revision 1, "Prairie Island Units I and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions," August 2013.
4 . LTR-LIS-14-50, "Prairie Island Units I and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction," January 2014.
Notes:
(a) 1l1is evaluation credits peaking factor bumdown, see Reference 2.
Page 17 of 36
Attachrnerifto LTR-LIS-16-40 February 18,2016 Page 4 of7.
Westinghouse LOCA Peak Clad Temperature Summa~-y for Appendix K Small Break Plant Name: Prairie Islarid Unit l Utility Name: Xcel Energy, Inc R,evision Date: 2/I/2016 Analysis Information EM: NO TRUMP Analysis Date: 1/21/2008 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 10 Notes: Zirlo ( l4X 14), Framatome RSG Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 959 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2015 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*
I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 959 It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.
References 1 . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units 1 and 2 PCT Summaries," February 2008.
Notes:
None Page 18 of36
Attachment to L TR-LIS-16-40 February 18,2016 Page 5 of7 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/112016 Ana.lysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Brcllk Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP(%): 10 Notes:
Clad Temp (DF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS
. Evaluation of Fuel Pellet Thennal Conductivity Degradation and Peaking 227 2 (a), (b)
Factor Bumdown 2 . Revised Heat Transfer Multiplier Distributions -2 3 3 . Error in Burst Strain Application 25 4 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2015 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*
I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2015 It is recommended that the licensee detennine if these PCTallocations should be considered with respect to I 0 CFR 50.46 reporting requirements.
References I . WCAP-17783-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Prairie Island Units I and 2 with Replacement Steam Generators Using ASTRUM Methodology," June 2013.
2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notification and Reporting tbr Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bum down, September 20, 2012.
3 . LTR-LIS-13-366, Revision I, "Prairie Island Units 1 and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions," August 2013.
- 4 . LTR-LIS-14-50, "Prairie Island Units I and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction," January 2014.
Notes:
(a) This evaluation credits peaking factor bumdown, see Reference 2.
(b) The reporting text and line item originally identified for Unit I in Reference 2 is applicable to Unit 2 with RSGs.
Page 19 of 36
Attachment to LTR-LIS-16-40 February 18,2016 Page 6 of7 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/112016 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Brealc Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 10 Notes: Zirlo (14Xl4), AREVA RSG Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 959 l, 2 a PCT ASSESSMENTS (DiMa PCT)
A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . None 0 C. 2015 ECCS MODEL ASSESSM!£NTS I . None 0 D. OTHER*
1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 959
- It is recommended that the licensee determine if these PCT allocations should be considered with respect to I0 CFR 50.46 reporting requirements.
References 1 . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units 1 and 2 PCT Summaries," February 2008.
2 . LTR-LlS-13-274, "Prairie fslalld Units I and 2, 10 CFR 50.46 Summary Sheets for the Evaluation to Support the Unil2 Installation of AREVA Model 56119 Replacement Steam Generators (RSGs)," June 2013.
Notes:
(a) The Unit I AOR is applicable to Unit 2 with the RSGs installed.
Page 20 of36
Attachment to L TR-LIS-16-40 February 18, 2016 Page 7 of7 10 CFR 50.46 Reporting SbarePoint Site Check:
EMs applicable to Prairie Island:
Realistic Large Break- ASTRUM (2004)
Appendix K Small Break- NOTRUMP 2015 Issnes Transmittal Letter Issue Description L TR-LIS-16-60 10 CFR 50.46 Report for Earlier Containment Spray Pump Actuation Times due to Degraded Fan Cooler Heat Removal Capacity Page 21 of 36