L-PI-07-057, Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

ML072360022
Person / Time
Site:	Prairie Island
Issue date:	08/23/2007
From:	Wadley M Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-07-057
Download: ML072360022 (15)

Text

Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC AUG 2 3 2007 L-PI-07-057 10 CFR 50.46 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie lsland Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Corrections to Emergencv Core Cooling System (ECCS) Evaluation Models Enclosed please find Attachment 1, "Westinghouse LOCA (loss of coolant accident)

Evaluation Model Changes, which is the 2006 annual report of corrections to the Prairie lsland Nuclear Generating Plant (PINGP) Units 1 and 2 ECCS Evaluation Models. This report is submitted in accordance with the provisions of 10 CFR 50.46 and summarizes changes made to both the large break LOCA (LBLOCA) and small break LOCA (SBLOCA) analyses.

The SBLOCA and LBLOCA Peak Clad Temperature (PCT) Assessment Sheets for Unit 1 and Unit 2 are enclosed as Attachment 2. The limiting LOCA analysis for Prairie lsland Unit 1 and Unit 2, with consideration of all 10 CFR 50.46 assessments, remains the LBLOCA analysis, as summarized in Attachment 2.

Neither Attachment 1 nor Attachment 2 need be withheld from public disclosure.

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

Michael D. Wadley v Site Vice president, Prairie lsland Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2 Attachments (2) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC

ATTACHMENT I NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS 50-282 AND 50-306 Westinghouse LOCA Evaluation Model Changes 3 Pages follow

Attachment 1 - 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-07-232 April 13,2007 Effect of a Fuel Reconstitution on the Prairie Island Unit 1 Peak Cladding Temperature (Non-Discretionary Change)

NOTE: THIS APPLIES TO UNIT 1 ONLY Backmound:

During Cycle 24 operation at Prairie Island Unit 1, it was determined that there was leaking fuel in one once-burned assembly. To mediate this situation, two new rods of natural uranium are replacing the leakers.

Affected Evaluation Modells)

Large Break LOCA SECY UP1 Evaluation Model Estimated Effect:

The Prairie Island Unit 1 Large Break LOCA SECY UP1 and Small Break LOCA NOTRUMP analyses of record were evaluated for a fuel reconstitution of two rods to establish the effect on PCT. The estimated effect is + 1O F on the PCT for both Large Break LOCA and Small Break LOCA.

Attachment 1 - 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-07-232 April 13,2007 GENERAL CODE MAINTENANCE (Enhancements/Forward-Fit Discretionary Change)

Back~round Various changes in code input and output format havc been made to cnhance usability and help preclude errors in analyses. This includes both input changes (e.g., more relevant input variables defined and more common input values used as defaults) and input diagnostics designed to preclude unreasonable values from being used, as well as various changes to code output which have no effect on calculated results. In addition, various updates were made to eliminate inactive coding, improve active coding, and enhance commenting, both for enhanced usability and to facilitate code debugging when necessary. These changes rcpresent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-1345 1.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated PCT impact of O°F.

Attachment 1 - 10 CFR 50.46 Reporting Text Our ref LTR-LIS-07-232 April 13,2007 Effect of a Fuel Reconstitution on the Prairie Island Unit 1 Peak Cladding Temperature (Non-Discretionary Change)

NOTE: THIS APPLIES TO UNIT 1 ONLY

Background:

During Cycle 24 operation at Prairie Island Unit 1, it was determined that there was leaking fuel in one once-burned assembly. To mediate this situation, two new rods of natural uranium are replacing the leakers.

Affected Evaluation Model(s1 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect:

The Prairie Island Unit 1 Large Break LOCA SECY UP1 and Small Break LOCA NOTRUMP analyses of record were evaluated for a fuel reconstitution of two rods to establish the effect on PCT. The estimated effect is + 1O F on the PCT for both Large Break LOCA and Small Break LOCA.

ATTACHMENT 2 NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS 50-282 AND 50-306 LBLOCA and SBLOCA Peak Clad Temperature Assessment Sheets 8 pages follow

Attachment 2 - PCT Rackup Sheets Our ref: LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie Island Unit 1 Utility Name: Nuclear Management Company, LLC Revision Date: 5/16/2006 Analysis Information EM: SECY UP1 Analysis Date: 3/1/1995 Limiting Break Size: Cd = 0.4 FQ: 2.4 FdH: 1.77 Fuel: OFA SGTP (%): 15 Notes: ZirloTM,OSG SGTP Evaluated up t o 24.64% (see also Note e); Fq increased to 2.5 (Item A.10); RSG Study at 10% SGTP.

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

PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I Fixed Heat Transfer Node Assignment Error/Accumulator Water -175 3 lnject~onError (1995 Report) 2 . I-D Trans~tionBoiling Heat Transfer Error (I997 Report) 59 5 3 . Vessel Channel DX Error (1 997 Report) -14 5 4 . Input Consistency (1 997 Report) 5 . No Items for 1996 & 1998 Reports 6 . Accumularor LineIPressurizer Surge Line Data 1 Plant Specific Accumulator Level & Line Volume / Plant Specific Restart Error:

Reanalysis (I 999 Repon) 7 . Modellng llpdates and Unheated Conductor Input Corrections (Plant Specific, 2000 Report) 8 . Accumulator Pressure +/- 30 psi Range (Plant Specific, 2001 Report) 9 . LHSl Error Evaluation (Plant Specific, 2002 Report) 30 13.14 (s) 10 . Sensit~vityStudy for FQ=2.5, LHSl Correction, etc. (as l~stedin note (f)) -47 16,18,19 (f,h)

(Plant Specific, 2003 Report) 1 I . Broken Loop Nozzle Loss Coefficient (Plant Specific) -19 18,19,21, (h) 25 12 . SECY Cold Leg Nozzle Expansion 13 25 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Sensitivity Study for Steam Generator Tube Plugging Increase to 25%

2 . Accumulator Water Volume +I- 25 ft3 Range I2 I1 3 . Accumulator Pressure Extended to +/- 55 psi Range 21 11 4 .2 Reconstituted Rods Evaluation 1 9 5 . SATP Core Average Bumup 17 20.22 6 . Sensit~vityStudy for Framatome Replacement Steam Generators 32 23 7 . HAUP LOCA Evaluation 3 24 C. 2007 ECCS MODEL ASSESSMENTS

Attachment 2 - PCT Rackup Sheets Our ref: LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie Island Unit 1 Utility Name: Nuclear Management Company, LLC Revision Date: 511 612006 I . None D. OTHER*

I . Removal of Reference 14 LHSl Error Evaluat~on LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 2043

• It is recommended that the licensee determine ~fthese PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I . 95NS-G-0021, "Updated UP1 LBLOCA," March 24, 1995 2 . WCAP-13919, Addendum I, "Prairie lsland Units 1 and 2 W C O B R m R A C Best Estimate UP1 Large Break LOCA Analysis Engineering Report Addendum 1: Updated Results," December 1996.

3 . NSP-96-202, "Northern States Power Company Prairie lsland Un~tsI and 2 10 CFR 50.46 Annual Notification and Reponing," February 20,1996.

4 . NSP-97-201, "Northern States Power Company Prair~eIsland Units I and 2 10 CFR 50.46 Annual Notification and Reporting," April 17, 1397.

5 . NSP-98-012, "Northern States Power Company Prairie lsland 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 Pra~rielsland Units I and 2 I0 CFR 50.46 Annual Notification and Reporting for 1998," April 29, 1999.

7 NSP-00-005, "Northern States Power Company Prairie lsland Units I and 2 10 CFR 50.46 Annual Notification and Reponing for 1999," February 2000.

8 . NSP-00-057, "Northern States Power Company Prairie lsland Un~ts1 and 2 LOCA Evaluation of 25% SGTP with Other Modeling Updates," December I 1. 2000.

9 . LTR-LIS-06-277, "Reconstitution Evaluation, 10 CFR 50.46 Reporting Plant Specific Text, and Updated Rackup Sheets for Prairie lsland Unit I, Cycle 24." 512006.

10 . NSP-01-006, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2000," March 6,2001.

1 I . NSP-02-9, "Nuclear Management Company Praine lsland Units I and 2 LBLOCA Accumulator Pressure and Volume Ranges Evaluation," February 15,2002.

12 . NSP-02-5, "Nuclear Management Company Prair~elsland Un~tsI and 2 10 CPK 50.46 Annual Not~ficationand Reporting for 2001," March 2002.

13 . NSP-02-59lLTR-ES1-02-194, "Final Evaluation of Large Break LOCA Error." December 2002 14 . NSP-03-19, "Nuclear Management Company Prairie lsland Un~tsI and 2 10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.

15 . MP92-TAH-0394 I ET-NSL-OPL-1-92-518, "NSPC Prairie lsland Unils 1 and 2. SG Tube Flaw Area Reduction under LOCA I SSE - Final Report". October 21, 1992.

16 . NSP-04-10 "Safety Analysis Transition Program Trans~nittalof Engineering Report," February 20, 2004 17 . NSP-93-513, Rev IIET-NSL-OPL-1-93-313, Rev. I, 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. 1 ", July 7, 1993. Confirmed by : Letter from K. E. Higar(NSP) to Mr. T. Hawley (W). "Acceptance ofNSP-93-513, Rev. I", July 30, 1993.

I8 . NSP-D4-38, "Nuclear Management Company Prairie lsland Units 1 and 2 I0 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.

19 . WCAP-16206-P, "SATP Engineering Report for Prairie Island," February 2004.

20 . NF-NMC-04-49, "Nuclear Management Company Prairie lsland Unit 1 Cycle 22 Final RSE," April 2004.

Attachment 2 - PCT Rackup Sheets Our ref LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie Island Unit 1 Utility Name: Nuclear Management Company, LLC Revision Date: 511 612006 21 . NSP-04-65, "Nuclear Management Company Prairie lsland Units 1 & 2 Safety Analysis Transition Program Repsonse to 10 CFR 50.46 Inquiry," April 21,2004.

22 . NF-NMC-04-129, "Nuclear Management Company Prairie lsland Unit I, Cycle 23 Final RSE," August 2004.

23 . NSP-04-114, "Nuclear Management Company Prairie lsland Units 1 & 2, Safety 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 Vessel Head Replacement Project, Prairie lsland Units 1 & 2," May 18,2005.

25 . NSP-05-19 1, "M~scellaneousLBLOCA SECY EM Error Notification," August 2005 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-01-030). NMC cognizant of uncertainty application and PCT sheet categorization (e) It is assumed that NMC is applying the 0.36% SGTP allowance factor branch of the SG LOCA / SSE issue (Reference 15).

Thus the 25% SGTP Study (Item B.l) supports a net SGTP limit of 24.64%.

(f) Sensitivity Study for. FQ=2.50, PAD 4.0 Implementation. Restoration of LHSl to Reference 17 values. SGlLoop AP Re-tuning, Core Power Restoration.

(g) The note (f) sensitivity study allows for the removal of the Reference 13 engineering assessment.

(h) Items A. 10 and A. I I presented as aggregate -66 OF entry prior to Reference 2 1 decomposition.

Attachment 2 - PCT Rackup Sheets Our ref LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 1 Utility Name: Nuclear Management Company, LLC Revision Date: 511 212006 Analysis Information EM: NOTRUMP Analysis Date: 11/21/2003 Limiting Break Size: 6 inch FQ: 2.8 FdH: 2 Fuel: OFA SGTP (%): 10 Notes: ZirloTM (14X14), Framatorne RSG Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None B. PLANNED PLANT MODIFICATION EVALUATIONS I . 2 Reconstituted Rods Evaluation C. 2007 ECCS MODEL ASSESSMENTS I . None D. OTHER*

I . None LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1410

• It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reprting requirements.

Rcfcrcnccs:

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

2 . WCAP-16206-P, "Safety Analysis Transition Program Engineering Report for the Prairie Island Nuclear Power Plant, Volume 1 Engineenng Analyses." February 2004.

3 . OC-PX-2004.009, "SBLOCA Analysis Loop Seal Restriction Option," Mercier to Brown, March 5, 2004.

4 . LTR-LlS-06-277, "Reconstitution Evaluation, 10 CFR 50.46 Reporting Plant Specific Text, and Updated Rackup Sheets for Prair~eIsland Unit I . Cycle 24." 512006.

Notes:

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

Attachment 2 - PCT Rackup Sheets Our refi LTR-LlS-07-232 April 13, 2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie Island Unit 2 Utility Name: Nuclear Management Company, LLC Revision Date: 2/23/2006 Analvsis Information EM: SECY UP1 Analysis Date: 3/1/1995 Limiting Break Size: Cd = 0.4 FQ: 2.4 FdH: 1.77 Fuel: OFA SCTP (%): 15 Notes: ZirloTM.SGTP Evaluated up to 24.64% (see also Note e); Fq increased to 2.5 (Item A.lO)

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

PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . Fixed Heat Transfer Node Assignment ErrorlAccurnulator Water Injection Error (1995 Report) 2 . I - D Transition Boiling Heat Transfer Error (1997 Report) 3 . Vessel Channel DX Error (1 997 Repor!)

4 . lnput Consistency ( 1997 Report) 5 . N o Items for 1996, 1998 & 2004 Reports 6 . Accumulator LinelPressurizer Surge Line Data IPlant Specific Accutnulator Level & Line Volume IPlant Specific Restart Error:

Reanalysis (1999 Report) 7 . Modeling Updates and Unheated Conductor lnput Corrections (plant spectfic) (2000 Report)

X . Accu~nulororPressure +I- 30 psi Range (Plant Specitic) (2001 Report) 9 . LHSl Error Evaluation (Plant Specific) (2002 Report) 10 . Sens~tivityStudy for FQ=2.5, LHSl Correction. etc. (as listed In note (t))

(Plan! Specific) (2003 Report)

II . Brokcn Loop Nozzle Loss Coefficient (Plant Specific) 12 S t C Y Cold Leg Nozzle Expansion B. PLANNED PLANT MODIFICATION EVALUATIONS I . Sensitivity Study for Steam Generator Tube Plugging Increase to 25'%

2 . Accu~nulatorWater Volun~e+I- 2 5 f13 Range 3 . Accu~nulatorPressure Extendcd to +I- 55 p s ~Range 4 . Cycle 22 SATP Core Average Burnup 5 . HAUP LOCA Evaluatton 6 . SATP Core Average Bumup Extension for Cycle 23 Kedes~g~l C. 2007 ECCS MODEL ASSESSMENTS 1 . None

Attachment 2 - PCT Rackup Sheets Our rep LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie lsland Unit 2 Utility Name: Nuclear Management Company, LLC Revision Date: 2/23/2006 D. OTHER*

I . Re~novalof Reference I 2 LHSl Error Evaluat~on LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2017

• It is recommended that the licensee determ~neif these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I . 95NS-'3-002 1. "Updated UP1 LBLOCA," March 24. 1995 2 . WCAP-13919, Addendum I."Prairie lsland Units 1 and 2 WCOBRNTRAC Best Estimate UP1 Large Break LOCA Analysis Engineering Report Addendum I: Updated Results." December 1996.

3 . NSP-96-202, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notiticalion and Reporting," February 20, 1996.

4 . NSP-97-201, "Northern States Power Company Prairie lsland Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting," April 17. 1997.

5 NSP-98-012, "Northern States Power Company Prairie lsland 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 lsland Units I 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 Not~ficationand Reporting for 1999." Febn~ary2000.

8 . NSP-00-057. "Northern States Power Company Prairie lsland Units 1 and 2 LOCA Evaluation o f 25% SGTP with Other Modeling Updates." December 1 1.2000.

9 . NSP-01-006. "Northern Statcs Power Company Praine lsland Units 1 and 2 I 0 CFR 50.46 Annual Notification and Reporting for 2000," March 6. 2001.

10 . NSP-02-9. "Nuclear Management Company Prairie lsland Units I and 2 LBLOCA Accumula~orPressure and Volume Ranges Evaluat~on."Februa~y15. 2002.

II . NSP-02-5. "Nuclear Managcment Company PI-airie lsland Units I and 2 10 CFR 50.46 Annual Notlticotion and Repolling for 2001." March 2002.

I2 NSP-02-59'LTR-ESI-02-lc)4."F~nalEvaluat~ono f Large Break LOCA ktuor," Decembel 2002 13 . NSP-03-19. "Nuclear Management Company Prairie lsland Units I and 2 10 CFR 50.46 Annual Notiticauon and Reporting for 2002." March 2003.

14 . MP92-TAH-0394 IEl-NSL-OPL-1-92-5 18. "NSPC Prairie lsland Units I and 2. SG Tubc Flow Area Reduction undcr LOCA !SSE - F~nalKcport", October 2 1. 1992.

15 . NSP-04-10 "Sat'cty Analysis Tranilnon Program Translnittal o f Enginecring Report," Feblua~y20, 2004.

16 . NSP-93-513, Rev IlKf-NSL-OPL-1-93-313, Rev. I.Letter from 1.. A. Hawley ( W ) to K. t. Higar(NSP). "Final Transmittal o f Assu~npt~ons to be used for the Large and Small Break LOCA Analyses. Rcv. I",July 7. 1993. Coniirmcd by : Letlcr from K. E. Higar (NSP) to Mr. T. Hawley (W). "Acceptance ofNSP-93-513. Rev. I".July 30. 1993.

17 . NSP-04-38. "Nuclear Management Company Prairie Island Units I and 2 10 CFR 50.46 Annual Notificntlon and Repon~ngfor 2003." March 2004.

18 . WCAP-16206-P. "SATP Engineering Report for Pralrie Island," February 2004.

19 . NF-NMC-04-50. "Nuclear Management Company Prairie lsland Unit 2 Cycle 22 F ~ n aUSE," l April 2004.

20 . NSP-04-65. "Nuclear Management Company Pra~rielsland Units I & 2 Safety Analysis Transition Program Repsonse to I 0 CFR 50.46 Inquiry." April 21.2004.

21 . NSP-05-155, "Nuclear Management Company. Reactor Vessel Head Replacement Project. Prair~eIsland Units I & 2," May

18. 2005.

Attachment 2 - PCT Rackup Sheets Our rep LTR-LIS-07-232 April 13,2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Large Break Plant Name: Prairie lsland Unit 2 Utility Name: Nuclear Management Company, LLC Revision Date: 2/23/2006 22 . NF-NMC-05-38 Rev. I, "Prairie lsland Unit 2 Cycle 23 Final RSE." May 13.2005 23 . NSP-05-65, "Nuclear Management Company Prairie lsland Units I and 2 10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

24 . NSP-05-191, "Miscellaneous LBLOCA SECY EM Error Notification." August 2005 Notes:

(a) P-bar-HA increased from I .57 to 1.59 (b) Reanalysis for all listed issues (c) Reanalysis for both issues (d) Related JCO in existence (NSP-0 1-030). NMC cognizant of uncertainty application and PCT sheet categorization.

(e) It is assumed that NMC is applylng the 0.36% SGTP allowance factor branch of the SG LOCA / SSE issue (Reference 14)

Thus the 25% SGTP Study (Item B.1) supports a net SGTP l~mltof 24.64%.

(f) Sensitivity Study for. FQ=2.50. PAD 4.0 Implementation. Restoration of LHSl to Reference 16 values, SGILoop AP Re-tuning. Core Power Restoration.

(g) The note (t) sensitivity study allows for the removal of the Reference 12 engineering assessment.

(11) Items A.10 and A.1 I presented as aggregate -66 "F entry prlor to Reference 20 decomposition

Attachment 2 - PCT Rackup Sheets Our ref LTR-LIS-07-232 April 13, 2007 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 2 Utility Name: Nuclear Management Company, LLC Revision Date: 2/23/2006 Analvsis Information EM: NOTRUMP Analysis Date: 9/1/2000 Limiting Break Size: 3 inch FQ: 2.8 FdH: 2 Fuel: OFA SGTP (%): 25 Notes: ZirloTM ( 1 4 x 1 4 )

Clad Temp ( O F ) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1142 I (a)

PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . No Items for 2000,2001 & 2002 Repons 0 2.4.5 2 . NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections 35 6.7 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None C. 2007 ECCS MODEL ASSESSMENTS I . None D. OTHER*

1 . Evaluation for Reduced Auxilary Feedwater Flow Rate LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1177

• It is recommended than the licensee determine if these PCT allocations be cor~s~det.cd with respect to I 0 CFR 50.46 reportlng requlrrrrlmts.

References:

I . NSP-00-045, "SBLOCA Re-analysis with Rev~sedNOTRUMP Code," Octoher 2,2000.

2 . NSP-01-006. "Northern States Power Company Prairie lsland Unlts 1 and 2 I 0 CFR 50.46 Annual Notification and Reporting for 2000." March 6. 2001.

3 . NSP-02-36, "SBLOC'A L.itnited FSAR Update and Evaluation for Revised Auxllery Feedu*aterFlow Ratc." October 2002.

4 . NSP-02-5. "Nuclcar Management Company Prairie lsland Units I and 2 10 CFR 50.46 Annual Notification and Reporting lor 2001 ." March 2002.

5 . NSP-03-19. "Nuclear Management Company Prairie lsland Unils I and 2 10 C F l l 50.46 Annual Not~ticationand Reporting for 2002." March 2003.

h . NSP-03-68. " I 0 CFR 50.46 Mid-Year Notification and Reporting fbr 2003." November 2003 7 . NSP-03-38. "Nuclear Management Company Prairie lsland Units I and 2 10 CFR 50.46 Annual Notificat~onand Reporting for 2003." March 2004.

Notes:

(a) Accumulator water volume sensitivity o f +/- 30 cubic feet included