L-PI-04-131, Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

ML043630311
Person / Time
Site:	Prairie Island
Issue date:	12/14/2004
From:	Solymossy J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-04-131
Download: ML043630311 (7)

Text

Commied Praie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC DEC 1 4 2004 L-PI-04-131 10 CFR 50.46 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Unit I Docket 50-282 License No. DPR-42 Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

Reference:

Letter L-PI-04-11, dated September 24, 2004, "Clarification of Actions for Corrections to Emergency Core Cooling System (ECCS) Evaluation Models," from Nuclear Management Company, LLC (NMC), to the Nuclear Regulatory Commission Attached is a report of changes to the Prairie Island Nuclear Generating Plant (PINGP)

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) and Small Break Loss of Coolant Accident (SBLOCA) changes reported by Westinghouse that are applicable from the beginning of Fuel Cycle 23.

The LBLOCA Peak Clad Temperature (PCT) changes are due to a sensitivity study performed for new steam generators (installed during the refueling outage that ended November 23, 2004). The change since the Last Acceptable Model submitted to the NRC on August 5, 2004 is +32 OF. The PCT (2026 IF, see Attachment 1) for the LBLOCA analysis continues to remain below the 10 CFR 50.46 PCT acceptance criterion. However, the accumulated absolute value of the PCT changes and errors since the original 1995 baseline Analysis of Record is 782 IF. NMC has committed by letter (see Reference) to provide a new LBLOCA analysis for PINGP by March 31, 2006.

The SBLOCA PCT changes are due to a reanalysis to address the installation of the new steam generators. The change since the Last Acceptable Model submitted to the NRC on August 5, 2004 is +232 IF. The PCT (1409 OF, see Attachment 2) for the SBLOCA 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 1717 Wakonade Drive East . Welch, Minnesota 55089-9642 Telephone: 651.388.1121 r

Document Control Desk Page 2 original baseline Analysis of Record is 267 IF. Since the current analysis was a full scope reanalysis, no schedule for reanalysis is necessary.

The summary sheets attached to this letter need not be withheld from public disclosure.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

Please contact Jack Leveille (651-388-1121, Ext 4142) if you have any questions related to this letter.

Vice Preside Pr ie Island Nuclear Generating Plant CC Regional Administrator, USNRC, Region IlIl Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector- Prairie Island Nuclear Generating Plant Enclosure

ENCLOSURE LOCA Peak Clad Temperature Summary Prairie Island Nuclear Generating Plant (includes plant specific changes and non-zero non-plant specific changes) 4 Pages Follow (Attachment 1 - Prairie Island Unit 1 LBLOCA- 3 pages)

(Attachment 2 - Prairie Island Unit 1 SBLOCA - 1 page)

A7rAcYMEt-7r Westinghouse LOCA Peak Clad Temperature Summary for SECY UPI Large Break Peoc 1k 3 Plant Name: Prairie Island Unit I Utility Name: Nuclear Management Company, LLC Revision Date: 8 /10104 Analysis Information EM: SECY UPI WCrT Analysis Date: 3/l/95 Lmiting Break Size: Cd = 0.4 FQ: 2.4 Fdll: 1.77 Fud: OFA SGTP (%): 15 Notes: ZirlorxI, OSG SGTP Evaluated up to 24.64% (see also Note n); Fq increased to 2.5 (Item A.10); RSG Study at 10% SGTP.

Clad Temp (°F) Ret. Notes LICENSING BASIS Analysis-Of-Record PCT 2180 1,2 (a)

MARGIN ALLOCATIONS (Delta PCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS I Fixed Heat Transfer Node Assignment Error/Accumubtor Water Injection -175 3 Error (1995 Report) 2 I -D Transition Boiling Heat Transfer Error (1997 Report) 59 5 3 Vessel Channel DX Error (1997 Report) -14 5 4 Input Consistency (1997 Report) .66 5 5 No Items for 1996. & 1998 Reports 0 4.6 6 Accumulator Line/Pressurizer Surge Line Data I Plant Specific 113 7 (b)

Accumublor Level & Line Volume I Plant Specific Restart Enor.

Reanalysis (I 999 Report)

7. Modeling Updates and Unheated Conductor Input Corrections (Plant -147 8,10 (c)

Specific. 2000 Report) 8 . Accumulator Pressure +I- 30 psi Range (Plant Specific. 2001 Report) S 12. 13 (d) 9 UISI Error Evaluation (Plant Specific. 2002 Report) 30 1415 (h) 10 Sensitivity Study for FQ-2.5, LHSI Correction. ctc. (as listed in note (g)) -47 17.19.20 (0~)

(Plant Specific. 2003 Report) 11 Broken Loop Nozzle Correction (Plant Specific) (2003 Report) -19 20.22 (i)

B. PLANNED PLANT CHANGE EVALUATIONS I Sensitivity Study for Steam Gcnerator Tube Plugging Incse to 25% 52 8

2. Accumulator Water Volume 1- 25 ft3 Range 12 12 3 . Accumulator Pressure Extended to +I- 55 psi Range 21 12 4 . 5 Reconstituted Rods Evaluation 0 9.11 (0)

5. SATP Core Average Burnup 17 21.23 6 . Sensitivity Study for farnatome RSG 32 24 C. 2004 PERMANENT ECCS MODEL ASSESSMENTS I. None 0 D. TEMPORARY ECCS MODEL ISSUES*

I None 0

A~TTA Cl- i '1.rI Westinghouse LOCA Peak Clad Temperature Summary for SECY UPI Large Break PI1f(01f 3.{3 Plant Name: Prairie Island Unit I Utility Name: Nuclear Management Company, LLC Revision Date: 8/10/04 E. OTHER I . Removal of Reference 14 LHSI Error Evaluation -30 17 (h)

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2026 It is recommended ihat these temporary PCT allocations which address current LOCA model issues no be considered with respect to 10 CFR 50.46 reporting requirements.

Rcrerences:

I . 95NS-G-0021. -Updated UPI LBLOCA. March 24. 1995.

2 . WCAP-I3919, Addendum 1, 'Prairie Island Units I and 2 WCOBRAJ7RAC Best Estimate UPI Large Break LOCA Analysis Engineering Report Addendum 1: Updated Results,' December 1996.

3 . NSP-96-202. 'Northern States Power Company Prairie bland Units I and 2 10 CFR 50.46 Annual Notification and Reporting.' February 20, 1996.

4 . NSP-97-201. "Northern States Power Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting.' April 17.1997.

5 . NSP-98-012. 'Northern States Power Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 1997,' February 27, 1998.

6 . NSP-99-010, "Northern States Power Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 1998.' April 29. 1999.

7 . NSP-005. Northem States Powe Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 1999,' February 2000.

8 . NSP-00-057, "Northern States Power Company Prairie Island Units I and 2 LOCA Evaluation of 25% SGTP with Other Modeling Updates,' December 11, 2000.

9. OONS-G-0076/CAB-00-390. 'Pr irie Island Unit I Cycle 21 LOCA Reload Confirmation and FCEP Checklist,' Decembcr IS.

2000.

10 . NSP-01-006. 'Northem States Powe Company Prairie Island Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for2000.' March6.2001.

11 . Rothrock (NMC) to Swigat (W). 'Prairie sland Unit I LOCA PCT. May 30.2001.

12 . NSP-02-9, 'Nuclear Management Company Prairie Island Units I and 2 LBLOCA Accumulator Pressure and Volume Ranges Evaluation.' February 15. 2002.

13 . NSP-02-5, "Nuclear Management Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 2001,' March 2002.

14. NSP-02-59/LTR-ESI-02-194. 'Final Evaluation o Large Break LOCA Error.' December 2002.

15 . NSP-03-19. 'Nuclear Management Company Prairie Iland Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 2002, March 2003.

16. MP92-TAti-0394 / ET-NSLOPL-I-92-5I8,'NSPC Prairie Island Units I and 2. SO Tube Flow Area Reduction under LOCAI SSE - Final Report'. October 21, 1992.

17 . NSP-04-IO "SafetyAnalysis Transition Progran Transmittal of Engineering Report," February 20.2004.

It . NSP-93-513, Rev IIET-NSLOPL 1-93-313. Rev. 1. Letter 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: Letter from K.

E. Higar (NSP) to Mr.T. Hawley (W), "Acceptance of NSP-93-5 13, Rev. 1'. July 30. 1993.

19 . NSP-04-38. 'Nuclear Management Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 2003.' March 2004.

20. WCAP-I 6206-P, "SATP Engineering Report for Prairie Island.' February 2004.

21 . NF-NMC.0449, "Nuclear Management Company Prairie Island Unit I Cycle 22 Final RSE." April 2004.

Westinghouse LOCA Peak Clad Temperature Summary for SECY UPI Large Break tP'ke 3 '(3 Plant Name: Prairie Island Unit I Utility Name: Nuclear Management Company, LLC Raevision Date: 8/10/04 22 NSP-04-65.Nuclear Management Company Prairie Island Units I & 2 Safety AnalysisTransition Program Repsonse to 10 CFR 50.46 Inquiry." April 21, 2004.

23 . F-NMC-04-xxx. Unit 1. Cycle 23 RSE. Aug. 2004.

24 . NSP-04-J 14, *Nuckar Management Company Prairie Island Units I & 2. Safety Analysis Transition Program, Transmittal of LBLOCA Replacement Stcam Generator (RSG) Engineering Report Addendum." (WCAP-16206-P-Addendum 1). June 2004.

Notes:

(a) P-bar-HA increased from 1.57 to 1.59 (b) Reanalysis for all listed issues (c) Reanalysis for both issues (d) Related JCO in existence (NSP-0N-030). NMC cognizant of uncertainty application and PCTshcet categorization.

(e) Reconstitution for Cycle 21 recanted per Reference 11.

(f) It is assumed that NMC is applying the 0.36% SGTP allowance factor branch of the SG LOCA / SSE issue (Reference 16).

Tbus the 25% SGTP Study (item B.I) supports a net SGTP limit of 24.64%.

(g) Sensitivity Study for: FQ2.50, PAD 4.0 Implementation. Restoration of LHSI to Reference I8 values. SG/Loop AP Re-tuning. Core Power Restoration.

(h) The note (g) sensitivity study allows for the removal of the Reference 14 engineering assessment.

(i) Items A.10 and A.l I presented as aggregate 466 *Fentry prior to Reference 22 decomposition.

che4- a p, , I.f I Westinghouse Proprietary Class 2 Attachment 2 - PCTr Sheets Page 7 of 10 Our ref: NSP-04-38 March 16,2004 Westinghouse LOCA Peak Clad Temperature Summary for Small Break Plant Name: Prairie Island Unit I Utility Name: Nuclear Management Company, LLC Future-B Revision Date: 3 /3/04 Analysis Infornation EMI: NOTRUMP Analysis Date: 11/21/03 Limiting Break Size: 6 inch FQ: 2.8 Fdll: 2 Fuel: OFA SGTP (%): 10 Notes: Zirlo'm(14X14), Framatome RSG Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1409 1,2 (a)

MARGIN ALLOCATIONS (Delta PCI)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS I .None 0 B. PLANNED PLANT CHANGE EVALUATIONS I .None 0 C. 2003 PERMANENT ECCS MODEL ASSESSMENTS I None 0 D. TEMPORARY ECCS MODEL ISSUES*

I . None 0 E. OTHER I .None 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1409 It is recommended that these temporary PCT allocations which address current LOCA model issues not be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I . NSP-04-10 'Safety Analysis Transition Program Transmittal of Engineering Report," February 20,2004.

2 . WCAP-16206-P, Safety AnalysisTransition Program Engineering Repont forthe Prairic Island Nudear Power Plant, Volume I Engineering Analyses," February 2004.

Notes:

(a) The 6-inch break is limiting when the loop seal restriction is applied to all break sizes.

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