ML060870393

Annual Operating Report of ECCS Model Changes

ML060870393
Person / Time
Site:	Wolf Creek
Issue date:	03/17/2006
From:	Moles K Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 06-0055
Download: ML060870393 (11)

Text

WOLF CREEK 'NUCLEAR OPERATING CORPORATION Kevin J. tVdoles March 17, 2006 Manager fegulatory Affairs RA 06-0055 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

Westinghouse Letter LTR-LIS-06-117, dated March 6, 2006, 10 CFR 50.46 Annual Notification and Reporting for 2005

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 2005 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 Report:ng." 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 aid 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 2005 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 2005. 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 2005. These model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rackup forms.

PO. Box 411 / Burlington, KS 66839/ Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNVET

RA 06-0055 Page 2 of 2 Attachment II provides the calculated Large Break Loss of Coolant Accident (LOCA) and Small Break L.OCA PCT margin allocations in effect for the 2005 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) 3644041.

KJM/rIt 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 W. B. Jones (NRC), w/a B. S. Mallet (NRC), w/a Senior Resident Inspector (NRC), w/a

Attachment I to RA 06-0055 Pace 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)

Noii-Discretionary Changes With PCT Impact None Non-Discretionary Changes With No PCT Impact Pressurizer Fluid Volumes Lower Guide Tube Assembly Weight Discrepancy in NOTRUMP RWST Draindown Calculation Enhancements/Forward-Fit Discretionary Changes General Code Maintenance (BASH/NOTRUMP)

Attachment I to RA 06-0055 Pace 2 of 3 PRESSURIZER FLUID VOLUMES

Background

The Westinghouse Systems and Equipment Engineering group has recommended that the previously-transmitted pressurizer fluid volumes be replaced with nominal cold values. This change resolves a discrepancy in the prior calculations while providing a close approximation of the actual as-built values. The revised values have been evaluated for impact on current licensing-basis analyses and will be incorporated into the plant-specific input databases on a forward-fit basis. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect I The differences between the previously-transmitted and revised volumes are very small and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 00F for 10 CFR 50.46 reporting purposes.

LOWER GUIDE TUBE ASSEMBLY WEIGHT

Background

An error was discovered in the lower guide tube assembly weight for three units that resulted in a small over-estimation of the upper plenum metal mass. The corrected values have been evaluated for impact on current licensing-basis analyses and will be incorporated into the plant-specific input databases on a forward-fit basis. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP.-

13451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in upper plenum metal mass are very small and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 00F for 10 CFR 50.46 reporting purposes.

Attachment I to RA 06-0055 Page 3 of 3 DISCREPANCY IN NOTRUMP RWST DRAINDOWN CALCULATION

Background

For small break LOCA calculations where the break size is greater than the safety injection (SI) line diameter, and where the SI line is connected directly to the reactor coo ant system (RCS), it is assumed that the broken loop safety injection flows do not inject to the RCS, but rather spill to containment. Typically, this is modeled in NOTRUMP-EM analyses by setting the flows injected to the broken loop equal to zero, which neglects the continued depletion of the refueling water storage tank (RWST) inventory. As a result, the RWST draindown time is incorrectly calculated, potentially resulting in an inaccurate modeling of enthalpy changes and/or SI interruptions that can occur at switchover to sump recirculation. Therefore, the SI spilling flows need to be explicitly modeled in order to correctly calculate the RWST draindown time.

Affected Evaluation Models 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect For Westinghouse plants using the NOTRUMP-EM, the larger small breaks are typically non-limiting and the transients are of short duration. Therefore, correct modeling of the spilling flows in the RWST draindown calculation for these breaks would be expected to produce a negligible effect on SBLOCA results, leading to an estimated PCT impact of 00F for 10 CFR 50.46 reporting purposes.

GE1NERAL CODE MAINTENANCE (BASH/NOTRUMP)

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-1 3451.

Affected Evaluation Models 19 1 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 00F.

Attachment 11 to RA 06-0055 Page 1 of 6

• • • 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=1.6 5 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 0F I (a)

MARGIN ALLOCATIONS (APCT)

A. PFRIOR 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. Containment Purge Evaluation 0 4

3. Cycle 10 Fuel Assembly Design Changes 95 5

4. Fuel Rod Crud 0 6 C. 2004 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 = 32IF SINCE LAST 30-DAY REPORT (LETTER ET 99-0045)

Attachment 11to RA 06-0055 Page 2 of 8

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, 19E97.

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 I ncrease," January 12, 1994.

5. Westinghouse to WCNOC letter 97SAP-G-0009, "Wolf Creek Nuclear Operating Corporation, Wolf (reek Generating Station, Safety Assessment for the Wolf Creek Generating Station with ZIRLO T Fuel Assemblies," February 7, 1997.

6. Westinghouse to WCNOC letter 97SAP-G-0075, "Wolf Creek Nuclear Operating Corporalion, 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, WNolf 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-00-1 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, 'Nolf 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, U10 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 < 2.50F was assessed and 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 06-0055 Page 3 of e5

• • • SMALL BREAK PEAK CLAD TEMPERATURE (PCT) MARGIN UTILIZATION Evaluation Model: 1985 EM with NOTRUMP Fuel: 17X17 V5H w/lFM, non-IFBA, 275 psig Peaking Factor: FQ=2.50, FdH=1. 65 SG Tube Plugging: 10%

Power Level: 3565 MWth1 Limiting transient: 3-inch Break LICENSING BASIS Clad Temp (OF) Ref. Notes Analy;ls of Record PCT 1510 1 MARGIN ALLOCATIONS (APCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS

1. Effect ofSI 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 Tracking/Region Depletion Errors 13 15

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

1. Loose Parts 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. 2004 PERMANENT ECCS MODEL ASSESSMENTS

1. None 0 D. TEMPORARY ECCS MODEL ISSUES

1. None 0 E. OTHER

1. Cold Leg Streaming Temperature Gradient 7 7

Attachment II to RA 06-0055 Page 4 of 6 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1672 CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES FIAPCTI = 350 F SINCE LAST 30-DAY REPORT (LETTER ET 99-0024)

Attachmen': II to RA 06-0055 Page 5 of 6

References:

1. Westinghouse Topical Report WCAP-13456, 'Wolf Creek Generating Station NSSS Reraling 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'm 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 Greek 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, 'Nolf 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, 'Nolf 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 Repcrting 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:

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

Attachment II to RA 06-0055 Page 6 of 13 (b) This ircrease 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.