ML051030295

From kanterella
Revision as of 04:07, 23 December 2019 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
CFR 50.46,Annual Report of ECCS Model Changes
ML051030295
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 04/07/2005
From: Moles K
Wolf Creek
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 05-0051
Download: ML051030295 (11)
v • d • e


Text

- WOLF CREEK 'NUCLEAR OPERATING CORPORATION Kevin J. Moles Manager Regulatory Affairs April 7, 2005 RA 05-0051 U. S. Nuclear Regulatory.Commission ATTN: Document Control Desk Washington, DC 20555

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 2004 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)(i), 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 2004. The review concludes that the effect of changes to, or errors in, the Evaluation Models on the limiting -transient PCT is not significant for 2004.

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 2004. These model changes and enhancements do not have impacts on the PCT and, generally, will not be presented on the PCT rackup forms.

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

RA 05-0051 Page2of2o Attachment II provides the calculated Large Break Loss of Coolant Accident (LOCA) and Small Break LOCA PCT margin allocations in effect for the 2004 WCGS evaluation models. Please

- note, that the PCT penalty associated with the high-head safety injection flow rate reduction was assessed as part of License Amendment No. -160, which changes Technical Specification - 3.5.5-1 to allow a higher seal injection flow. 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 requireeients 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.

- - truly yours,

~ ~~Very --

.Kevin J. Moles KJM/rlg

-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 (POT) Margin Utilization cc: J. N. Donohew (NRC), w/a D. N. Graves (NRC), w/a B. S. Mallet (NRC), w/a.

Senior Resident Inspector (NRC), w -a

Attachment I to RA 05-0051 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 PCT Impact LOCBART Fluid Property Logic Non-Discretionary Changes With No PCT Impact Steam Generator Inlet/Outlet Plenum Flow Areas LBLOCA Initial Containment Relative Humidity Assumption Enhancements/Forward-Fit Discretionary Changes General Code Maintenance (BASH/NOTRUMP)

Attachment I to RA 05-0051 Page 2 of 3 LOCKBART FLUID PROPERTY LOGIC

Background

Several minor discrepancies related to the LOCBART fluid property logic were discovered and corrected. For example, the routine used to calculate the enthalpy and specific volume of superheated steam was renamed to resolve a naming conflict with a library routine that uses different logic to calculate the same parameters. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effect Representative plant calculations using the LOCBART code generally showed either no effect or a negligible effect on results, with some tendency for a small increase in peak cladding temperature (PCT) for plants with an early-reflood PCT. For these plant categories, 10 CFR 50.46 assessments were developed either to bound the representative plant calculations or on a plant-specific basis.

STEAM GENERATOR INLET/OUTLET PLENUM FLOW AREAS

Background

The basis for calculating the steam generator inlet and outlet plenum flow areas used with the SATAN-VI momentum flux model has been redefined as the average area over the plenum height. This change resolves a discrepancy in the original calculation and provides a more appropriate basis for the corresponding flow area terms. This change represents a Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effect Calculations using the SATAN-VI code indicated that this change has a negligible effect on the blowdown thermal-hydraulic transient results that will be assigned a 0F PCT impact for 10 CFR 50.46 reporting purposes.

Attachment I to RA 05-0051 Page 3 of 3 LBLOCA INITIAL CONTAINMENT RELATIVE HUMIDITY ASSUMPTION

Background

Large break LOCA analyses have historically used maximum initial relative humidity to specify the initial containment air and steam partial pressures. This assumption is conservative for a given total initial containment pressure, but is non-conservative for a given initial containment air partial pressure. The historical assumption has been revised to reflect this distinction, and the analysis input guidelines have been updated accordingly. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effect An evaluation for the plants within Westinghouse Pittsburgh large break LOCA analysis cognizance concluded that no PCT assessments are required, leading to an estimated PCT effect of 00F for 10 CFR 50.46 reporting purposes.

GENERAL 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 comments, 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 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 0F.

Attachment II to RA 05-0051 i 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 (OF) Ref. Notes Analysis of Record PCT 19160 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 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 0F CUMULATIVE ABSOLUTE MAGNITUDE OF PCT CHANGES 7-IAPCTI = 32 0F SINCE LAST 30-DAY REPORT (LETTER ET 99-0045)

Attachment II to RA 05-0051 Page2of6

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 ZIRLOm 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-00-1 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.

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 05-0051 X Page3of6

      • SMALL BREAK 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 (OF) 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 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 4 (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 05-0051

  • 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)

Attachment II to RA 05-0051 Page 5 of 6

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 ZIRLOm 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-00-118, "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.
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:

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

Attachment II to RA 05-0051 Page 6 of 6 (b) This increase in auxiliary feedwater temperature was originally evaluated in Reference 8 as a 160F 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.

Retrieved from "https://kanterella.com/w/index.php?title=ML051030295&oldid=2288762"