ML090760983

CFR 50.46 Annual Report for Emergency Core Cooling System (ECCS) Model Changes for 2008

ML090760983
Person / Time
Site:	Wolf Creek
Issue date:	03/13/2009
From:	Flannigan R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 09-0062
Download: ML090760983 (8)

Text

WELF CREEK WNUCLEAR OPERATING CORPORATION Richard D. Flannigan March 13, 2009 Manager Regulatory Affairs RA 09-0062 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

Westinghouse Letter LTR-LIS-09-93, dated January 29, 2009, Wolf Creek 10 CFR 50.46 Annual Notification and Reporting for 2008

Subject:

Docket No. 50-482: 10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Gentlemen:

This letter provides the annual report for the Emergency Core Cooling System (ECCS)Evaluation Model changes and errors for the 2008 model year that affect the Peak Cladding Temperature (PCT) for Wolf Creek Generating Station (WCGS). This letter is provided in accordance with the criteria and reporting requirements of 10 CFR 50.46(a)(3)(ii), as clarified in Section 5.1 of WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." Regulation 10 CFR 50.46(a)(3)(ii) states, in part, "For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in section 50.4. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with section 50.46 requirements." Wolf Creek Nuclear Operating Corporation (WCNOC) has reviewed the notification of 10 CFR 50.46 reporting information pertaining to the ECCS Evaluation Model changes that were implemented by Westinghouse for 2008 as described in the above Reference.

The review concludes that the effect of changes to, or errors in, the Evaluation Models on the limiting transient PCT is not significant for 2008. Therefore, the report of the ECCS Evaluation Model changes is provided on an annual basis.Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2008. These model changes and enhancements do not have impacts on the PCT and, generally, will not belpresented on the PCT rackup forms.P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET RA 09-0062 Page 2 of 2 Attachment II provides the calculated Large Break Loss of Coolant Accident (LOCA) and Small Break LOCA PCT margin allocations in effect for the 2008 WCGS evaluation models. The PCT values determined in the Small Break and Large Break LOCA analysis of record, combined with all of the PCT allocations, remain well below the 10 CFR 50.46 regulatory limit of 2200 degrees Fahrenheit.

Therefore, WCGS is in compliance with 10 CFR 50.46 requirements and no reanalysis or other action is required.No commitments are identified in this correspondence.

If you have any questions concerning this matter, please contact me at (620) 364-4117, or Diane Hooper at (620) 364-4041.Sincerely Richard D. Flannig RDF/rlt Attachment I -Attachment II -Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss of Coolant Accidents (LOCA)Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization cc: E. E. Collins (NRC), w/a V. G. Gaddy (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a Attachment I to RA 09-0062 Page 1 of 3 ASSESSMENT OF CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS FOR LARGE AND SMALL BREAK LOSS OF COOLANT ACCIDENTS (LOCA)Non-Discretionary Changes With Peak Cladding Temperature (PCT) Impact None Non-Discretionary Changes With No PCT Impact Errors in Reactor Vessel Lower Plenum Surface Area Calculations (BASH/NOTRUMP)

Discrepancy in Metal Masses Used from Drawings (BASH/NOTRUMP)

Enhancements/Forward-Fit Discretionary Changes General Code Maintenance (BASH/NOTRUMP)

Editorial Changes Editorial Changes to Wolf Creek (SAP) Small Break Loss of Coolant Accident (LOCA)PCT Rackup Sheets Attachment I to RA 09-0062 Page 2 of 3 Errors in Reactor Vessel Lower Plenum Suiface Area Calculations (Non-Discretionary Changes with no PCT Impact)Background Two errors were discovered in the calculations of reactor vessel lower plenum surface area. The corrected values have been evaluated for impact on current licensing-basis analysis results and will be incorporated on a forward-fit basis. 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)1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in vessel lower plenum surface area are relatively minor and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.Discrepancy in Metal Masses Used from Drawings (Non-Discretionary Changes with no PCT Impact)Background Discrepancies were discovered in the use of metal masses from drawings.

The updated reactor vessel metal masses and fluid volumes have been evaluated for impact on current licensing-basis analysis results and will be incorporated on a forward-fit basis.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)1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in the reactor vessel metal mass and fluid volume are relatively minor and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

Attachment I to RA 09-0062 Page 3 of 3 General Code Maintenance (Enhancements/Forward-Fit Discretionary Changes)Background Various changes have been made to enhance the usability of the codes and to help preclude errors in analyses.

This includes items such as 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-1 3451.Affected Evaluation Model(s)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 0°F.Editorial Changes to Wolf Creek (SAP) Small Break LOCA PCT Rackup Sheets Background The purpose of this report is to document various editorial changes to Wolf Creek (SAP)Small Break LOCA PCT Rackup sheets. The old Small Break LOCA PCT Rackup Sheets that were based on the analysis in Reference 1 are no longer applicable and are to be removed. The new Rackup sheet that was based on the analysis of record (AOR)for Wolf Creek (Reference

2) contains "Cycle 17" indication.

This indication is to be deleted as the new Rackup sheet is also applicable to cycles following Cycle 17 until further notice.Affected Evaluation Model(s)N/A (see note in Estimated Effect section)Estimated Effect Since these changes are editorial, they do not have any-effect on the results of the current licensing analysis.Reference(s)

1. WCAP-13456, "Wolf Creek Generating Station NSSS Rerating Licensing Report", October 1992.2. WCAP-16717-P, Rev. 0, "Wolf Creek Generating Statinon (SAP), MSIV/MFIV Replacement Project, Small Break Loss of Coolant Accident Analysis Engineering Report", January 2007.

Attachment II to RA 09-0062 Page 1 of 3 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION

• LARGE BREAK LOCA PEAK CLAD TEMPERATURE (PCT) MARGIN UTILIZATION Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting transient:

LICENSING BASIS 1981 EM with BASH 17x17 V5H w/IFM, non-IFBA, 275 psig FQ=2.50, FdH=1.65 10%3565 MWth Cd=0.4, Min. SI, Reduced Tavg Clad Temp (°F)1916°F Analysis of Record PCT Ref. Notes 1 (a)MARGIN ALLOCATIONS (APCT)A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. Structural Metal Heat Modeling 2. LUCIFER Error Corrections

3. Skewed Power Shape Penalty 4. Hot Leg Nozzle Gap Benefit 5. SATAN-LOCTA Fluid Error 6. LOCBART Spacer Grid Single-Phase Heat Transfer Error 7. LOCBART Vapor Film Flow Regime Heat Transfer Error 8. LOCBART Cladding Emissivity Errors 9. LOCBART Radiation to Liquid Logic Error Correction

10. LOCBART Pellet Volumetric Heat Generation Rate B. PLANNED PLANT CHANGE EVALUATIONS

1. Loose Parts Evaluation

2. Effects of Containment Purging 3. Cycle 10 Fuel Assembly Design Changes 4. Fuel Rod Crud C. 2008 PERMANENT ECCS MODEL ASSESSMENTS

1. None D. TEMPORARY ECCS MODEL ISSUES E. OTHER 1. Cold Leg Streaming Temperature Gradient 2. Rebaseline of AOR (12/96)3. LOCBART Zirc-Water Oxidation Error-25-6 152-136 15 15 9 6 17 45 20 0 95 0 8 10 11 11 2 9 12 13 14 15 3 4 5 6 0 0 0-63 28 8 9 7 (b)(c)(d)LICENSING BASIS PCT + MARGIN ALLOCATIONS CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES SINCE LAST 30-DAY REPORT (LETTER ET 07-0021)PCT = 2088 0 F E I APCTI = 0F Attachment II to RA 09-0062 Page 2 of 3

References:

1. Westinghouse Topical Report WCAP-13456, "Wolf Creek Generating Station NSSS Rerating Licensing Report," October 1992.2. Westinghouse to WCNOC letter SAP-97-102, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Annual Notification and Reporting," February 17, 1997.3. Westinghouse to WCNOC letter SAP-90-148, "Wolf Creek Nuclear Operating Corporation, RCS Loose Parts Evaluation," April 18, 1998.4. Westinghouse to WCNOC letter SAP-94-102, "Containment Mini Purge Isolation Valve Stroke Time Increase," January 12, 1994.5. Westinghouse to WCNOC letter 97SAP-G-0009, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Safety Assessment for the Wolf Creek Generating Station with ZIRLO T M Fuel Assemblies," February 7, 1997.6. Westinghouse to WCNOC letter 97SAP-G-0075, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Wolf Creek Crud Deposition/Axial Offset Anomaly Safety Evaluation," September 29, 1997.7. Westinghouse to WCNOC letter OOSAP-G-0006, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Wolf Creek Cycle 12 LOCA Current Limits," February 10, 2000.8. Westinghouse to WCNOC letter SAP-93-701, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Notification and Reporting Information," January 25, 1993.9. Westinghouse to WCNOC letter SAP-99-148, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 BART/BASH Evaluation Model, Mid-Year Notification and Reporting for 1999," September 22, 1999.10. Westinghouse to WCNOC letter SAP-94-703, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Notification and Reporting Information," February 8, 1994.11. Westinghouse to WCNOC letter SAP-95-716, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, LOCA Axial Power Shape Sensitivity Model," August 14, 1995.12. Westinghouse to WCNOC letter SAP-00-1 18, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Appendix K (BART/BASH/NOTRUMP)

Evaluation Model, Mid-Year Notification and Reporting for 2000," June 30, 2000.13. Westinghouse to WCNOC letter SAP-00-150, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 BART/BASH Evaluation Model Mid-Year Notification and Reporting for 2000," December 2000.14. Westinghouse to WCNOC letter SAP-02-32, "10 CFR 50.46 BART/BASH Evaluation Model Mid-Year Notification and Reporting for 2002," June 2002.15. LTR-LIS-07-312, "10 CFR 50.46 Reporting Text for LOCBART Version 37.0 Issues and Revised PCT Rackup Sheets for Wolf Creek," May 2007.Notes: (a) An evaluation was performed to support removal of the transition core penalty for Cycle 12 (Ref. 7).(b) A PCT benefit of < 2.5°F was assessed, however, a benefit of 0°F will be tracked for reporting purposes.(c) This previously unclaimed benefit was realized through prior rebaseline of the limiting case.(d) This assessment is a function of analysis PCT plus certain margin allocations and as such may increase/decrease with margin allocation changes.

Attachment II to RA 09-0062 Page 3 of 3***SMALL BREAK LOCA PEAK CLAD TEMPERATURE (PCT) MARGIN UTILIZATION

• • • Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting transient:

LICENSING BASIS 1985 EM with NOTRUMP 17x17 RFA-2 w/IFM FQ=2.50, FdH=1.65 10%3565 MWth 4-inch Break Clad Temp (TF)936 Ref. -Notes 1 Analysis of Record PCT MARGIN ALLOCATIONS (APCT)A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. None B. PLANNED PLANT CHANGE EVALUATIONS

1. Loose Part Evaluation C. 2008 PERMANENT ECCS MODEL ASSESSMENTS

1. None D. TEMPORARY ECCS MODEL ISSUES 1. None E. OTHER 1. None LICENSING BASIS PCT + MARGIN ALLOCATIONS 0 45 2 (a)0 0 0 PCT = 981°F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES EI APCT = 0-F

References:

1. WCAP-16717-P, Rev. 0, "Wolf Creek Generating Station (SAP), MSIV/MFIV Replacement Project, Small Break Loss of Coolant Accident Analysis Engineering Report," January 2007.2. SAP-90-148/NS-OPLS-OPL-I-90-239, "Wolf Creek Nuclear Operating Corporation, RCS Loose Part Evaluation," April 1990.Notes: (a) This penalty will be carried to track the loose part which has not been recovered.