ML070860756

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

ML070860756
Person / Time
Site:	Wolf Creek
Issue date:	03/16/2007
From:	Moles K Wolf Creek
To:	Document Control Desk, NRC/NRR/ADRO
References
LTR-LIS-07-148, RA 07-0032
Download: ML070860756 (11)

Text

WF CREEK NUCLEAR OPERATING CORPORATION Kevin J. Moles March 16, 2007 Manager Regulatory Affairs RA 07-0032 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

Westinghouse Letter LTR-LIS-07-148, dated March 7, 2007, 10 CFR 50.46 Annual Notification and Reporting for 2006 - BASH-EM and NOTRUMP-EM

Subject:

Docket No. 50-482: 10 CFR 50.46 Annual Report of ECCS Model Changes Gentlemen:

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

Evaluation Model changes and errors for the 2006 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 above reference that addresses 10 CFR 50.46 reporting information pertaining to the ECCS Evaluation Model changes that were implemented by Westinghouse for 2006. The review concludes that the effect of changes to, or errors in, the Evaluation Models on the limiting transient PCT is not significant for 2006. 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 2006. These model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rack-up forms.

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 4t1)0 An Equal Opportunity Employer M/F/HC/VET

RA 07-0032 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 2006 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) 364-4126, or Ms.

Diane Hooper at (620) 364-4041.

Sincerely, KJM/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: J. N. Donohew (NRC), w/a V. G. Gaddy (NRC), w/a B. S. Mallet (NRC), w/a Senior Resident Inspector (NRC), w/a

Attachment I to RA 07-0032 Page 1 of 4 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 BASH Minimum and Maximum Time Step Sizes PAD Version 4.0 Implementation NOTRUMP-EM Refined Break Spectrum EnhancementslForward-Fit Discretionary Changes General Code Maintenance (BASH/NOTRUMP)

Attachment I to RA 07-0032 Page 2 of 4 BASH Minimum and Maximum Time Step Sizes (Non-Discretionary Changes With No PCT Impact)

Background

A review of some recent BASH-EM sensitivity calculations led to a recommendation to reduce the minimum and maximum time step sizes in BASH during reflood. These changes are being recommended for generic application and have been evaluated for impact on existing analysis results. 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 Appendix K Large Break LOCA Evaluation Model with BASH Estimated Effect Sensitivity calculations using BASH and SMUUTH show that reducing the minimum and maximum time step sizes in BASH during reflood results in either a negligible change or a modest increase in the integral flooding rate for most cases, leading to an estimated impact of 0°F for 10 CFR 50.46 reporting purposes. One case showed a decrease in the integral flooding rate late in reflood and was evaluated for 10 CFR 50.46 impact on a plant-specific basis. For Wolf Creek Generating Station (WCGS), plant specific assessments conclude that implementation of these corrections leads to a conservative estimate of 00 F PCT effect for 10 CFR 50.46 reporting purposes.

PAD Version 4.0 Implementation (Non-Discretionary Changes With No PCT Impact)

Background

A recent BASH-EM evaluation predicted an increase in the peak cladding temperature (PCT) for IFBA fuel that was attributed primarily to the use of fuel rod initial conditions based on PAD Version 4.0. This result called into question the basis for forward-fit implementation of PAD Version 4.0, and existing IFBA analyses based on PAD Version 3.4 were reviewed to identify conditions that could lead to similar behavior. For each potentially-affected analysis, the estimated PCT impact due to PAD Version 4.0 implementation was assessed on a plant-specific basis. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451.

Affected Evaluation Model(s) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH

Attachment I to RA 07-0032 Page 3 of 4 Estimated Effect The 10 CFR 50.46 assessments for this issue were determined on a plant-specific basis, and for WCGS a 00 F PCT effect was assessed for 10 CFR 50.46 reporting purposes.

NOTRUMP-EM Refined Break Spectrum (Non-Discretionary Changes With No PCT Impact)

Background

During the course of reviewing several extended power uprate and replacement steam generator Small Break LOCA (SBLOCA) analyses, the Nuclear Regulatory Commission (NRC) questioned the break spectrum analyzed in the NOTRUMP evaluation model (EM). The NRC was concerned that the resolution of the break spectrum used in the NOTRUMP EM (1.5, 2, 3, 4, and 6 inch cases) may not be fine enough to capture the worst break with regard to limiting peak clad temperature as per 10 CFR 50.46. That is, the plant could be SBLOCA limited with regard to overall LOCA results.

In response to this, Westinghouse performed some preliminary work indicating that in some cases more limiting results could be obtained from non-integer break sizes; however, the magnitude of the impact was far less than that shown in preliminary work performed by the NRC. Based on this, Westinghouse performed evaluations to determine if all currently operating plants would maintain compliance with the 10 CFR 50.46 acceptance criteria when considering a refined SBLOCA break spectrum. It should be noted that use of a refined break spectrum is not an error, but a change, since evaluating only integer break sizes has been the standard practice since the initial licensing of NOTRUMP.

Affected Evaluation Model(s) 1985 Appendix K Small Break LOCA Evaluation Model Estimated Effect Consistent with the method described in Reference 1, for plants with low SBLOCA peak cladding temperatures (PCTs) (i.e., less than 1700 0 F) and overall SBLOCA results that are significantly non-limiting when compared with large break LOCA (LBLOCA) results, no explicit refined break spectrum calculations were performed, leading to an estimated impact of 00 F for 10 CFR 50.46 reporting purposes. For plants with high SBLOCA PCTs (i.e., equal to or greater than 1700 0 F), explicit refined break spectrum calculations were performed, and PCT penalties were assessed, if necessary.

Reference

1. LTR-NRC-06-44, "Transmittal of LTR-NRC-06-44 NP-Attachment, 'Response to NRC Request for Additional Information on the Analyzed Break Spectrum for the Small Break Loss of Coolant Accident (SBLOCA) NOTRUMP Evaluation Model (NOTRUMP EM),

Revision 1,' (Non-Proprietary)," July 14, 2006.

Attachment I to RA 07-0032 Page 4 of 4 GENERAL CODE MAINTENANCE (BASH/NOTRUMP)

(Enhancements/Forward-Fit Discretionary Changes)

Background

Various changes in code input and output format have been made to enhance 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 represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 1981 Appendix K Large Break LOCA Evaluation Model with BASH 1985 Appendix K Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated PCT impact of 0°F.

Attachment II to RA 07-0032 Page 1 of 5 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION

• • • LARGE BREAK LOCA PEAK CLAD TEMPERATURE (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 (OF) Ref. Notes Analysis of Record PCT 1916°F I (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 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. 2006 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 = 2043°F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES EIAPCTI = 320 F SINCE LAST 30-DAY REPORT (LETTER ET 99-0045)

Attachment II to RA 07-0032 Page 2 of 5

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 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," 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-00-118, "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.

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.

.4¸ *!, * :* * ,. *

• Attachment II to RA 07-0032 Page 3 of 5 SMALL BREAK LOCA PEAK CLAD TEMPERATURE (PCT) MARGIN UTILIZATION Evaluation Model: 1985 EM with NOTRUMP 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: 3-inch Break LICENSING BASIS Clad Temp (*F) Ref. Notes Analysis of Record PCT 1510 1 MARGIN ALLOCATIONS (APCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. Effect of SI in Broken Loop 150 10

2. Effect of Improved Condensation Model -150 10

3. Drift Flux Flow Regime Errors -13 11

4. LUCIFER Error Corrections -16 11

5. Boiling Heat Transfer Correlation Error -6 12

6. Steam Line Isolation Logic Error 18 12

7. Axial Nodalization, RIP Model Revision and SBLOCTA Error 26 13 Corrections Analysis

8. NOTRUMP Specific Enthalpy Error 20 2

9. SBLOCTA Fuel Rod Initialization Error 2 14

10. NOTRUMP Mixture Level TrackinglRegion Depletion Errors 13 15

11. NOTRUMP Bubble Rise/Drift Flux Model Inconsistency 0 16 Corrections B. PLANNED PLANT CHANGE EVALUATIONS

1. Loose Part Evaluation 45 3

2. Cycle 10 Fuel Assembly Design Change 1 6

3. Reduced Feedwater Inlet Temperature 10 4

4. Fuel Rod Crud 4 5 (a)

5. Auxiliary Feedwater Temperature Increase 16 8,9 (b)

6. High Head SI Flow Reduction 35 17 C. 2006 PERMANENT ECCS MODEL ASSESSMENTS

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

1. Cold Leg Streaming Temperature Gradient 7 7 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1672°F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES E IAPCT = 35°F SINCE LAST 30-DAY REPORT (LETTER ET 99-0024)

Attachment II to RA 07-0032 Page 4 of 5

References:

1. Westinghouse Topical Report WCAP-13456, "Wolf Creek Generating Station NSSS Rerating Licensing Report," October 1992.,

2. Westinghouse to WCNOC letter SAP-96-705, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Notification and Reporting," February 9, 1996.

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-96-119, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Small Break LOCA Evaluation for Reduced Feedwater Temperature," May 30,1996.

5. 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. (This penalty will be carried until such time it is determined to no longer apply).

6. 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.

7. 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.

8. Westinghouse to WCNOC letter SAP-98-138, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Assessment of an Increase in Auxiliary Feedwater Temperature,"

July 23, 1998.

9. 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.

10. Westinghouse to WCNOC letter SAP-93-718, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Safety Injection in the Broken Loop," September 22, 1993.

11. Westinghouse to WCNOC letter SAP-94-703, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Notification and Reporting," February 8, 1994.

12. Westinghouse to WCNOC letter SAP-94-722, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, 10 CFR 50.46 Notification and Reporting," August 18, 1994.

13. Westinghouse to WCNOC letter SAP-94-727, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, SBLOCTA Axial Nodalization," October 27, 1994.

14. 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.

15. 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.

16. Westinghouse to WCNOC letter SAP-03-33, "10 CFR 50.46 Mid-Year Notification and Reporting for 2003," November 14, 2003.

17. Westinghouse to WCNOC letter SAP-04-33, "Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, High Head Safety Injection Flow Rate Reduction -Final Evaluation,"

June 11,2004.

Notes are on the next page.

Attachment II to RA 07-0032 Page 5 of 5 Notes:

(a) This penalty will be carried until such time it is determined to no longer apply.

(b) This increase in auxiliary feedwater temperature was originally evaluated in Reference 8 as a 160 F penalty. However, this change was not implemented until the Cycle 12 reload. Reference 9 represents the transmittal of the Cycle 12 LOCA Reload Current Limits.