ML12100A083

Submittal of Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes

ML12100A083
Person / Time
Site:	Wolf Creek
Issue date:	03/27/2012
From:	Sen G Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 12-0028, LTR-LIS-12-109
Download: ML12100A083 (8)

Text

W0LF CREEK 'NUCLEAR OPERATING CORPORATION Gautam Sen Manager Regulatory Affairs March 27, 2012 RA 12-0028 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

Westinghouse Letter LTR-LIS-12-109, dated February 28, 2012, "Wolf Creek 10 CFR 50.46 Annual Notification and Reporting for 2011"

Subject:

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

This letter provides the annual report for the Emergency Core Cooling System (ECCS)

Evaluation Model changes and errors for the 2011 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 holder of a construction permit, operating license, combined license, or manufacturing license 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 §50.4 or §52.3 of this chapter, as applicable. 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 §50.46 requirements."

Wolf Creek Nuclear Operating Corporation (WCNOC) has reviewed the notification and reporting requirements of 10 CFR 50.46 pertaining to the ECCS Evaluation Model changes that were implemented by Westinghouse for 2011 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 2011. Therefore, changes to the ECCS Evaluation Model are being reported as an annual report.

Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2011. These model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rackup forms.

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer MIF/HCNET

RA 12-0028 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 2011 WCGS evaluation models. The PCT values determined in the Large Break and Small Break LOCA analysis of record, combined with all of the PCT allocations, remain well below the 10 CFR 50.46(b)(1) regulatory limit of 2200 OF.

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) 364-4175, or William Muilenburg at (620) 364-8831, Extension 4511.

Sincerely, Gautam Sen GS/rlt Attachment I - Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss of Coolant Accidents (LOCA)

Attachment II - Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization cc: E. E. Collins (NRC), w/a J. R. Hall (NRC), w/a N. F. O'Keefe (NRC), w/a Senior Resident Inspector (NRC), w/a

Attachment I to RA 12-0028 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 Radiation Heat Transfer Logic (NOTRUMP)

Maximum Fuel Rod Time Step Logic (NOTRUMP)

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

Editorial Changes None

Attachment I to RA 12-0028 Page 2 of 3 Summary RADIATION HEAT TRANSFER LOGIC (Non-Discretionary Changes with no PCT Impact)

Background

Two errors were discovered in the calculation of the radiation heat transfer coefficient in the SBLOCTA computer code. First, existing diagnostics did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated. These calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1 degree. Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine. These errors have been corrected in the SBLOCTA code and represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting."

Affected Evaluation Model(s) 1985 Westinghouse Small Break Loss of Coolant Accident (LOCA) Evaluation Model with NOTRUMP Estimated Effect A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of 00 F for existing Small Break LOCA analysis results.

MAXIMUM FUEL ROD TIME STEP LOGIC (Non-Discretionary Changes with no PCT Impact)

Background

An error was discovered in the SBLOCTA code that allowed the fuel rod time step to exceed the specified maximum allowable time step. The time step logic has been corrected in the SBLOCTA code. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of 00 F for existing Small Break LOCA analysis results.

Attachment I to RA 12-0028 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-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 nature of these changes leads to an estimated PCT impact of 00 F.

Attachment II to RA 12-0028 Page 1 of 3 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION LARGE BREAK LOCA PCT MARGIN UTILIZATION Evaluation Model: 1981 EM with BASH Fuel: 17x17 V5H w/IFM, non-IFBA, 275 psig Peaking Factor: FQ=2.50, FdH=1.65 SG Tube Plugging: 10%

Power Level: 3565 MWth Limiting transient: Cd=0.4, Min. SI, Reduced Tavg LICENSING BASIS Clad Temp (°F) 'Ref. Notes Analysis of Record PCT 1916°F 1 (a)

MARGIN ALLOCATIONS (APCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. Structural Metal Heat Modeling -25 8

2. LUCIFER Error Corrections -6 10

3. Skewed Power Shape Penalty 152 11

4. Hot Leg Nozzle Gap Benefit -136 11

5. SATAN-LOCTA Fluid Error 15 2

6. LOCBART Spacer Grid Single-Phase Heat Transfer Error 15 9

7. LOCBART Vapor Film Flow Regime Heat Transfer Error 9 12

8. LOCBART Cladding Emissivity Errors 6 13

9. LOCBART Radiation to Liquid Logic Error Correction 17 14

10. LOCBART Pellet Volumetric Heat Generation Rate 45 15 B. PLANNED PLANT CHANGE EVALUATIONS

1. Loose Parts Evaluation 20 3

2. Effects of Containment Purging 0 4

3. Cycle 10 Fuel Assembly Design Changes 95 5

4. Fuel Rod Crud 0 6 C. 2011 PERMANENT ECCS MODEL ASSESSMENTS

1. None 0 D. TEMPORARY ECCS MODEL ISSUES 0 E. OTHER

1. Cold Leg Streaming Temperature Gradient 0 8 (b)

2. Rebaseline of AOR (12/96) -63 9 (c)

3. LOCBART Zirc-Water Oxidation Error 28 7 (d)

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2088°F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES EIAPCTI =0°F SINCE LAST 30-DAY REPORT (LETTER ET 07-0021)

Attachment II to RA 12-0028 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, 1990.

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 TM 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-00-1 50, "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 0 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 12-0028 Page 3 of 3 SMALL BREAK LOCA PCT MARGIN UTILIZATION ***

Evaluation Model: 1985 EM with NOTRUMP Fuel: 17x17 RFA-2 w/IFM Peaking Factor: FQ=2.50, FdH=1.65 SG Tube Plugging: 10%

Power Level: 3565 MWth Limiting transient: 4-inch Break LICENSING BASIS Clad Temp (°F) Ref. Notes Analysis of Record PCT 936 1 MARGIN ALLOCATIONS (APCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. None 0 B. PLANNED PLANT CHANGE EVALUATIONS

1. Loose Part Evaluation 45 2 (a)

C. 2011 PERMANENT ECCS MODEL ASSESSMENTS

1. None 0 D. TEMPORARY ECCS MODEL ISSUES

1. None 0 E. OTHER

1. None 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 981°F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES E JAPCTI = 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.