L-PI-13-053, Annual Report of Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

Annual Report of Corrections to Emergency Core Cooling System (ECCS) Evaluation Models

ML13179A040
Person / Time
Site:	Prairie Island
Issue date:	06/26/2013
From:	Jeffery Lynch Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-13-053
Download: ML13179A040 (34)

Text

Xcel Energy J~N 2 6. 2013 L-PI-13-053 10 CFR 50.46 l.JJ.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 Renewed License Nos. DPR-42 and DPR-60 2012 Annual Report of Corrections to the Prairie Island Nuclear Generating Plant (PINGP) Emergency Core Cooling System (ECCS) Evaluation Models Pursuant to 10 CFR 50.46, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM") submits the 2012 annual report of corrections to the PINGP Units 1 and 2 ECCS evaluation models.

Attachment 1 contains the "Non-Plant Specific LOCA Errors and Changes" and summarizes the changes made to both the large break LOCA (LBLOCA) and small break LOCA (SBLOCA) analysis.

The limiting LOCA analysis peak clad temperature (PCT) for PINGP Unit 1 and Unit 2, with consideration of all 10 CFR 50.46 assessments, remains the LBLOCA analysis as summarized in Attachment 2.

A historical summary of the LOCA PCT for both Prairie Island Unit 1 and Unit 2 changes and errors to the LBLOCA and SBLOCA Evaluation Models are included in Attachment

2. Attachment 3 contains the 50.46 PCT Rack-up sheets. Attachment 4 is the Westinghouse Letter LTR-LIS-13-1 01 revision 1.

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

<L~t:::

Site Vice-President, Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota 1717 Wakonade Drive East

• Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2 Attachments (4) cc: Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC

ATTACI-IMENT 1 Non-Plant Specific LOCA Errors and Changes (See the proceeding 9 pages)

Attachment to LTR-LIS-13-101, Revision I March 15,2013 Page 1 of14 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, tmits 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) 1996 Westinghouse Best Esthnate Large Break LOCA Evaluation Model 1999 Westinghouse Best Esthnate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM

• Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) hnpact of0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 2 of 14 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN

Background

Fuel pellet thermal conductivity degradation (TCD) and peaking factor bumdown were not explicitly considered in the Prairie Island Unit 1 Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 (Reference 1) notified addressees of recent infonnation obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted peak cladding temperature (PCT) results in realistic emergency core cooling system (ECCS) evaluation models. This evaluation provides an estimated effect of fuel pellet TCD and peaking factor bumdown on the PCT calculation for the Prairie Island Unit 1 BE LBLOCA AOR. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451 (Reference 2).

Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation, as discussed in Reference 3, was performed to assess the PCT effect of fuel pellet TCD and peaking factor burndown on the Prairie Island Unit 1 BE LBLOCA analysis and concluded that the estimated PCT impact is 227°F for 10 CPR 50.46 reporting purposes. The peaking factor bumdown, included in the evaluation, is provided in Table 1 and is conservative for the current cycle. Xcel Energy, Inc. and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.

T a ble 1 : P ea ki ng F actors A ssume d.m th e E va uahon ofTCD Rod Burnup FdH <1).(2> FQ Transient <1> FQ Steady-State (MWd/MTU) 0 1.770 2.500 2.250 30,000 1.770 2.500 2.250 60,000 1.400 1.889 1.700 62 000 1.400 1.889 1.700 (1) Includes uncertamtles.

(2) Hot assembly average power follows the same burndown, since it is a function ofFdH.

References

1. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011. (NRC ADAMS# ML113430785)

2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CPR 50.46 Reporting,"

October 1992.

3. OG-12-386, "For Information Only- Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0) (Proprietary/Non-Proprietary)," September 18, 2012.

Attachment to LTR-LIS-13-101; Revision 1 March 15, 2013 Page 3 of 14 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN Bacl{ground Fuel pellet thennal conductivity degradation (TCD) and peaking factor bumdown were not explicitly considered in the Prairie Island Unit 2 Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 (Reference 1) notified addressees of recent information obtained concerning the impact of

.irradiation on fuel thennal conductivity and its potential to cause significantly higher predicted peak cladding temperature (PCT) results in realistic emergency core cooling system (ECCS) evaluation models. This evaluation provides an estimated effect of fuel pellet TCD and peaking factor burndown on the PCT calculation for the Prairie Island Unit 2 BE LBLOCA AOR. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451 (Reference 2).

Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation, as discussed in Reference 3, was performed to assess the PCT effect of fuel pellet TCD and peaking factor burndown on the Prairie Island Unit 2 BE LBLOCA analysis and concluded that the estimated PCT impact is 340°F for 10 CFR 50.46 reporting purposes. The peaking factor burndown, included in the evaluation, is provided in Table 1 and is conservative for the current cycle. Xcel Energy, Inc. and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.

ang F ac tors Assume dm' t h T a ble 1 : P eaI' eEvaluat i on ofTCD Rod Burnup FdH <1M2> FQ Transient <1> FQ Steady-State

{MWd/MTU) 0 1.770 2.500 2.250 30,000 1.770 2.500 2.250 60,000 1.400 1.889 1.700 62,000 1.400 1.889 1.700 (1) Includes uncertainties.

(2) Hot assembly average power follows the same bumdown, since it is a function ofFdH.

• References

1. NRC Infonnation Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thennal Conductivity Degradation," December 13, 2011. (NRC ADAMS# ML113430785)

2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

3. OG-12-386, "For Infonnation Only- Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0) (Proprietary/Non-Proprietary)," September 18, 2012.

Attachment to LTR-LIS-13-101, Revision I March 15, 2013 Page 4 of 14 HOTSPOT BURST TEMPERATU RE CALCULATION FOR ZIRLO CLADDING

Background

1 A problem was identified in the calculation of the burst temperature for ZIRL0 cladding in the

• HOTSPOT code when the cladding engineering hoop stress exceeds 15,622 psi. This problem results in either program failure or an invalid extrapolation of the burst temperature vs. engineering hoop stress table. This problem has been evaluated for impact on existing analyses, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The evaluation of existing analyses demonstrated no impact on the overall Peak Cladding Temperature (PCT) results, leading to an estimated effect of 0°F.

1 ZIRLO is a registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States ofAmerica and may be registered in other countries. All rights reserved. Unauthorized use is strictly prohibited.

Attachment to LTR-LIS-13-1 01, Revision 1 March 15, 2013 Page 5 of 14 HOTSPOT ITERATION ALGORITHM FOR CALCULATING THE INITIAL FUEL PELLET AVERAGE TEMPERATURE

Background

The HOTSPOT code has been updated to incorporate the following corrections to the iteration algorithm for calculating the initial fuel pellet average temperature: (1) bypass the iteration when the input value satisfies the acceptance criterion; (2) prevent low-end extrapolation of the gap heat transfer coefficient; (3) prevent premature tennination of the iteration that occurred under certain conditions; and (4) prevent further adjustment of the gap heat transfer coefficient after reaching the iteration limit. These changes represent a closely-related group ofNon-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1'996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using A STRUM Estimated Effect Sample calculations and engineering judgment lead to an estimated Peak Cladding Temperature (PCT) impact of0°F.

Attachment to LTR-LIS-13-1 0 I, Revision I March 15, 2013 Page 6 of 14 WCOBRA/TRAC AUTOMATED RESTART PROCESS LOGIC ERROR

Background

A minor error was identified in the WCOBRA/TRAC Automated Restart Process (WARP) logic for defining the Double-Ended Guillotine (DEG) break tables. The error has been evaluated for impact. on current licensing-basis analysis results and will be incorporated into the plant-specific analyses on a forward-fit basis. These changes represent a closely-related group ofNon-Discretionary Changes in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect These errors were evaluated to have a negligible impact on the Large Break LOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of0°F.

Attachment to LTR-LIS-13-101, Revision I March 15, 2013 Page 7 of 14 ROD INTERNAL PRESSURE CALCULATION

Background

Several issues which affect the calculation of rod internal pressure (RJP) have been identified for certain Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) evaluation models (EMs). These issues include the sampling of rod internal pressure uncertainties, updating HOTSPOT to consider the effect of transient RlP variations in the application of the uncertainty, and generating RlPs at a consistent rod power. These issues have been evaluated to estimate the impact on existing LBLOCA analysis results. The resolution of these issues represents a closely-related group ofNon-Discretionary Changes in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The effects described above are either judged to have a negligible effect on existing LBLOCA analysis results or have been adequately incorporated into the thermal conductivity degradation evaluations, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 8 of 14 WCOBRA/TRAC THERMAL-HYDRAULIC IDSTORY FILE DIMENSION USED IN HSDRIVER

Background

A problem was identified in the dimension of the WCOBRAJTRAC thermal-hydraulic history file used in HSDRJVER. The array that is used to store the information from the WCOBRAJTRAC thermal-hydraulic history file is dimensioned to 3000 in HSDRJVER. It is possible for this file to contain more than 3000 curves. If that is the case, it is possible that the curves would not be used correctly in the downstream HOTSPOT execution. An extent-of-condition review indicated that resolution of this issue does not impact the Peak Cladding Temperature (PCT) calculation for prior Large Break Loss-of-Coolant Accident (LBLOCA) analyses. This represents a Discretionary Change in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect As discussed in the Background section, resolution of this issue does not impact the PCT calculation for prior LBLOCA analyses, which leads to a PCT impact of 0°F.

Attachment to LTR-LIS-13-1 01, Revision 1 March 15, 2013 Page 9 of 14 NOTRUMP-EM EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION

Background

An evaluation has been completed to estimate the effect of fuel pellet thennal conductivity degradation (TCD) on peak cladding temperature (PCT) for plants in the United States with analyses using the 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (NOTRUMP-EM). This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) .

1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Based on the phenomena and physics of the SBLOCA transient, in combination with limited sensitivity calculations, it is concluded that TCD has a negligible effect on the limiting cladding temperature transient, leading to an estimated PCT impact of0°F.

ATTACHMENT 2 PCT HISTORICAL

SUMMARY

NOTE: This is a summary of the LOCA PCT as of 12/31/12. It does not include any changes effective in 2013 or any other anticipated changes.

PllLBLOCA PllSBLOCA PI2LBLOCA PI2SBLOCA Analysis of 11/30/07 1/21/08 11/30/07 1/21/08 Record Date Analysis of 1765 959 1623 965 Record PCT Old Changes 0 0 0 0 and Errors (absolute sum)

Last NRC 9/16/12 (2) 6/28/11 (1) 09/16/12 (2) 6/28/11 (1)

Notification Date Projected PCT 1992 959 1963 965 From the Last NRC Notification New Errors or 0 0 0 0 Changes (from this Notification Only)

Total of Changes 227 0 340 0 and Errors (absolute Sum)

New Projected 1992 959 1963 965 PCT (1) These dates reflect the SBLOCA 2011 annual report L-PI-12-046.

(2) These dates reflect the 30 day submittal issued for TCD in report L-PI-12-095

ATTACHMENT 3 LOCA Peak Clad Temperature Summary (Rack-Up Sheets)

Prairie Island Nuclear Generating Plant (includes plant specific changes and non-zero non-plant specific changes)

(See the proceeding 4 pages)

Attachment to LTR-LIS- I 3- I 0 I, Revision I March I 5, 2013 Page 10 of 14 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Breal<

Plant Name: Prairie Island Unit I Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Breal{ Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 10 Notes:

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2012 ECCS MODEL ASSESSMENTS I . Evaluation of Fuel Pellet Thennal Conductivity Degradation and Peaking 227 2 (a)

Factor Burndown D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1992

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

References I . WCAP-16890-P, Revision I, "Best-Estimate Analysis ofthe Large-Break Loss-of-Coolant Accident for the Prairie Island Nuclear Plant Unit I Using ASTRUM Methodology," June 2008.

2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20,2012.

Notes:

(a) This evaluation credits peaking factor bumdown, see Reference 2.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 11 of 14 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 1 Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel:* 422 Vantage+ SGTP (%): 10 Notes: Zirlo (14Xl4), Framatome RSG Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 959 1 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2012 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 I . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units I and 2 PCT Summaries," February 2008.

Notes:

None

Attachment to LTR-LIS-13-101, Revision l March 15, 2013 Page 12 of 14 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: 2/28/2013 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 25 Notes:

Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1623 PCT ASSESSMENTS {Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2012 ECCS MODEL ASSESSMENTS I , Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking 340 2 (a)

Factor Bumdown D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1963

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

References 1 . WCAP-16891-P, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Prairie Island Nuclear Plant Unit 2 Using ASTRUM Methodology," June 2008.

2 . LTR-LIS-12-414, "Prairie Island Units l and 2, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown," September 20, 2012.

Notes:

(a) This evaluation credits peaking factor bumdown, see Reference 2.

Attachment to LTR-LIS-13-101, Revision I March 15, 2013 Page 13 of 14 Westinghouse LOCA Pea){ Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Break Size: 2 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 25 Notes: Zirlo (14X14)

Clad Temp eF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 965 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2012 ECCS MODEL ASSESSMENTS 1 . None 0 D. OTHER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 965

• 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 I . LTR-LIS-08-158, "Transmittal of Future Prairie Island Unils I and 2 PCT Summaries," February 2008.

Notes:

None

ATTACHMENT 4 WESTINGHOUSE LETTER LTR-LIS~l3-101 revision 1 (Reference 1 of this letter)

NOTE: This letter was retrieved from the Westinghouse "E-Room" using internet access.

NOTE: This letter includes non-plant specific errors and changes.

(See the proceeding 15 pages)

Westinghouse NoncProprietary Class 3 8 Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-6663 Direct fax: (724) 720-0857 e-mail: utleyd@westinghouse.com Our ref: LTR-LIS-13-101, Revision 1 March 15, 2013 Prairie Island Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2012

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 2012 model year. Standardized reporting pages for all changes and errors for the Evaluation Models utilized for your plant(s) are enclosed, consistent with the commitment following the NUPIC audit in early 1999. 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 of 0°F will generally not be presented on the PCT sheet. Any plant-specific errors in the application of the model for 2012 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 of the 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 infonnation that is provided in this letter was prepared in accordance with Westinghouse's Quality Management System (QMS).

Tlte changes to Revision 1 oftlzis letter are limited to the removal of the PCT rackup sheets associated with the Extended Power Uprate program.

Author: (Electronically Approved)* Verified: (Electronically Approved)*

Dania! W. Utley Rosemary T. Null LOCA Integrated Services II LOCA Integrated Services II Approved: (Electronically Approved)*

Dawn M. Crytzer Manager, LOCA Integrated Services II

Attachment:

10 CFR 50.46 Reporting Text and PCT Summary Sheets (14 Pages)

• Electronically approved records are attthen/icated in the electronic document management system.

© 2013 Westinghouse Electric Company LLC All Rights Reserved

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 1 of 14 GENERALCODEN VUNTENANCE

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 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with*

Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 2 of 14 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN Bacl{ground

,Fuel pellet thermal conductivity degradation (TCD) and pealdng factor burndown were not explicitly

.considered in the Prairie Island Unit 1 Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 (Reference 1) notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted peak cladding temperature (PCT) results in realistic emergency core cooling system (ECCS) evahJ.ation models. This evaluation provides an estimated effect of fuel pellet TCD and pealdng factor burndown on the PCT calculation for the Prairie Island Unit 1 BE LBLOCA AOR. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451 (Reference 2).

Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation, as discussed in Reference 3, was performed to assess the PCT effect of fuel pellet TCD and pealdng factor burndown on the Prairie Island Unit 1 BE LBLOCA analysis and concluded that the estimated PCT impact is 227°F for 10 CFR 50.46 reporting purposes. The peaking factor burndown, included in the evaluation, is provided in Table 1 and is conservative for the current cycle. Xcel Energy, Inc. and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.

on2 F actors A ssume d'm th e E va uatwn ofTCD T a bl e 1 : P ea I' RodBurnup FdH <1>*<2> FQ Transient <1> FQ Steady-State (MWd/MTU) 0 1.770 2.500 2.250 30,000 1.770 2.500 2.250 60,000 1.400 1.889 1.700 62,000 1.400 1.889 1.700 (1) Includes uncertainties.

(2) Hot assembly average power follows the same burndown, since it is a function ofFdH.

References

1. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L.A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13,2011. (NRC ADAMS# ML113430785)

2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

3. OG-12-386, "For Information Only- Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0) (Proprietary/Non-Proprietary)," September 18, 2012.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 3 of14 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN

Background

Fuel pellet thennal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the Prairie Island Unit 2 Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 (Reference 1) notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted peak cladding temperature (PCT) results in realistic emergency core cooling system (ECCS) evaluation models. This evaluation provides an estimated effect of fuel pellet TCD and peaking factor burndown on the PCT calculation for the Prairie Island Unit 2 BE LBLOCA AOR. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451 (Reference 2).

Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A quantitative evaluation, as discussed in Reference 3, was performed to assess the PCT effect of fuel pellet TCD and peaking factor burndown on the Prairie Island Unit 2 BE LBLOCA analysis and concluded that the estimated PCT impact is 340°F for 10 CFR 50.46 reporting purposes. The pealdng factor burndown, included in the evaluation, is provided in Table 1 and is conservative for the current cycle. Xcel Energy, Inc. and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.

ang F actors A ssume d.m t h e E va uatton ofTCD T a ble 1 : P ea I' Rod Burnup FdH (t),(z) FQ Transient <*> FQ Steady-State (MWd/MTU) 0 1.770 2.500 2.250 30,000 1.770 2.500 2.250 60,000 1.400 1.889 1.700 62,000 1.400 1.889 1.700 (1) Includes uncertainties.

(2) Hot assembly average power follows the same burndown, since it is a function ofFdH.

References

1. NRC Information Notice 2011-21, McGinty; T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011. (NRC ADAMS # ML113430785)

2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

3. OG-12-386, "For Information Only- Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0) (Proprietary/Non-Proprietary)," September 18, 2012.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 4 of 14 HOTSPOT BURST TEMPERATURE CALCULATION FOR ZIRLO CLADDING

Background

1 A problem was identified in the calculation of the burst temperature for ZIRL0 cladding in the HOTSPOT code when the cladding engineering hoop stress exceeds 15,622 psi. This problem results in either program failure or an invalid extrapolation of the burst temperature vs. engineering hoop stress table. This problem has been evaluated for impact on existing analyses, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.

AffectedEvaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 .Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The evaluation of existing analyses demonstrated no impact on the overall Peak Cladding Temperature (PCT) results, leading to an estimated effect of0°F.

1 ZIRLO is a registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States ofAmerica and may be registered in other countries. All rights reserved. Unauthorized use is strictly prohibited.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 5 of 14 HOTSPOT ITERATION ALGORITHM FOR CALCULATING THE INITIAL FUEL PELLET AVERAGE TEMPERATURE

Background

The HOTSPOT code has been updated to incorporate the following corrections to the iteration algorithm for calculating the initial fuel pellet average temperature: (1) bypass the iteration when the input value satisfies the acceptance criterion; (2) prevent low-end extrapolation of the gap heat transfer coefficient; (3) prevent premature termination of the iteration that occurred under certain conditions; and (4) prevent further adjustment of the gap heat transfer coefficient after reaching the iteration limit. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Sample calculations and engineering judgment lead to an estimated Peak Cladding Temperature (PCT) impact of0°F.

Attachment to LTR-LIS-13-1 01, Revision 1 March 15, 2013 Page 6 of 14 WCOBRA/TRAC AUTOMATED RESTART PROCESS LOGIC ERROR

Background

A minor error was identified in the WCOBRA/TRAC Automated Restart Process (WARP) logic for defining the .Double-Ended Guillotine (DEG) break tables. The error has been evaluated for impact on current licensing-basis analysis results and will be incorporated into the plant-specific analyses on a forward-fit basis. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation ModelUsing ASTRUM Estimated Effect These errors were evaluated to have a negligible impact on the Large Break LOCA analysis results, leading to an estimated Peak Cladding Temperature (PCT) impact of0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 7 of 14 ROD INTERNAL PRESSURE CALCULATION

Background

Several issues which affect the calculation of rod internal pressure (RIP) have been identified for certain Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) evaluation models (EMs). These

• • issues include the sampling of rod internal pressure uncertainties, updating HOTSPOT to consider the effect of transient RIP variations in the application of the uncertainty, and generating RIPs at a consistent rod power. These issues have been evaluated to estimate the impact on existing LBLOCA analysis results. The resolution ofthese issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The effects described above are either judged to have a negligible effect on existing LBLOCA analysis results or have been adequately incorporated into the thermal conductivity degradation evaluations, leading to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 8 of 14 WCOBRA/TRAC THERMAL-HYDRAULIC IDSTORY FILE DIMENSION USED IN HSDRIVER Bacl<ground A problem was identified in the dimension of the WCOBRA/TRAC thermal-hydraulic history file used in HSDRJVER. The array that is used to store the information from the WCOBRA/TRAC thermal-hydraulic history file' is dimensioned to 3000 in HSDRJVER. It is possible for this file to contain more than 3000 curves. If that is the case, it is possible that the curves would not be used correctly in the downstream HOTSPOT execution. An extent~of-condition review indicated that resolution of tlris issue does not impact the Peak Cladding Temperature (PCT) calculation for prior Large Break Loss-of-Coolant Accident (LBLOCA) analyses. This represents a Discretionary Change in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect As discussed in the Background section, resolution of this issue does not impact the PCT calculation for prior LBLOCA analyses, which leads to a PCT impact of0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 9 of 14 NOTRUMP-EM EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION

Background

An evaluation has been completed to estimate the effect of fuel pellet thennal conductivity degradation (TCD) on peak cladding temperature (PCT) for plants in the United States with analyses using the 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (NOTRUMP-EM). This change represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Based on the phenomena and physics of the SBLOCA transient, in combination with limited sensitivity calculations, it is concluded that TCD has a negligible effect on the limiting cladding temperature transient, leading to an estimated PCT impact of0°F.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 10 of 14 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: 2/28/2013 Analysis Information EM: ASTRUM (2004) Analysis Date: ll/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP(%): 10 Notes:

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1765 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS 1 . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2012 ECCS MODEL ASSESSMENTS I . Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking 227 2 (a)

Factor Bumdown D. OTHER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1992

• It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CPR 50.46 reporting requirements.

References I . WCAP-16890-P, Revision I, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Prairie Island Nuclear Plant Unit I Using ASTRUM Methodology," June 2008.

2 . LTR-LIS-12-414, "Prairie Island Units 1 and 2, 10 CPR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown," September 20,2012.

Notes:

(a) This evaluation credits peaking factor bumdown, see Reference 2.

Attachment to LTR-LIS-13-101, Revision 1 March 15,2013 Page 11 of 14 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 1 Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Brealt Size: 3 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 10 Notes: Zirlo (14X14}, 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. 2012 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 ifthese PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References I . LTR-LIS-08-158, "Transmittal of Future Prairie Island Units 1 and 2 PCT Summaries," February 2008.

Notes:

None

Attachment to LTR-LIS-13-101, Revision I March 15, 2013 Page 12 of 14 Westinghouse LOCA Peak Clad Temper ature Summar y for ASTRUM Best Estimate Large Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Informat ion EM: ASTRUM (2004) Analysis Date: 11/30/2007 Limiting Break Size: Split FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 25 Notes:

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1623 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0

B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . None 0

C. 2012 ECCS MODEL ASSESSMENTS 1 . Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking 340 2 (a)

Factor Burndown D. OTHER*

I . None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1963 It is recommended that the licensee determine ifthese PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References I . WCAP-16891-P, "Best-Estimate Analysis of the Large-Break Loss-of-Co olant Accident for the Prairie Island Nuclear Plant Unit 2 Using ASTRUM Methodology," June 2008.

2 . LTR-LIS-12-414, "Prairie Island Units I and 2, 10 CFR 50.46 Notificatio n and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20, 2012.

Notes:

(a) This evaluation credits peaking factor bumdown, see Reference 2.

Attachment to LTR-LIS-13-101, Revision 1 March 15, 2013 Page 13 of 14 Westinghous e LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Prairie Island Unit 2 Utility Name: Xcel Energy, Inc Revision Date: 2/28/2013 Analysis Information EM: NOTRUMP Analysis Date: 1/21/2008 Limiting Break Size: 2 inch FQ: 2.5 FdH: 1.77 Fuel: 422 Vantage+ SGTP (%): 25 Notes: Zirlo (14X14)

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 965 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMEN TS 1 . None 0 B. PLANNED PLANT MODIFICAT ION EVALUATIONS 1 . None 0 C. 2012 ECCS MODEL ASSESSMEN TS 1 . None 0 D. OTHER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 965

" 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 I and 2 PCT Summaries," February 2008.

Notes:

None

Attachment to LTR-LIS-13-101, Revision i March 15, 2013 Page 14 of 14 10 CFR 50.46 Reporting SharePoin t Site Check:

EMs applicable to Prairie Island:

Realistic Large Break- ASTRUM (2004)

Appendix K Small Break- NOTRUM P 2012 Issues Transmitt al Letter Issue Descriptio n L TR-LIS-12-3 87 10 CFR 50.46 Report for HOTSPOT Burst Temperatu re Calculatio n for ZIRLO Cladding LTR-LIS-12-414 Prairie Island Units 1 and 2 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductiv ity Degradation and Peaking Factor Burndown L TR-LIS-12-482 10 CFR 50.46 Report for the Rod Internal Pressure Calculation LTR-LIS-12-555 10 CFR 50.46 Report for WCOBRA JTRAC Thermal-H ydraulic Risto_!}' File Dimension Used in HSDRIVE R LTR-LIS-12-602 10 CFR 50.46 Report for HOTSPOT Iteration Algorithm for Calculating the Initial Fuel Pellet Average Temperatu re