ML19086A109
| ML19086A109 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/18/2019 |
| From: | Benham R Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA 19-0034 | |
| Download: ML19086A109 (9) | |
Text
,.,,
Ron Benham March 18, 2019 Manager Nuclear and Regulatory Affairs RA 19-0034 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Reference:
Westinghouse Letter LTR-LIS-19-33, dated February 13, 2019, "Wolf Creek 10 CFR 50.46 Annual Notification and Reporting for 2018"
Subject:
Docket No. 50-482: 10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes To Whom It May Concern:
In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the annual reporting requirement for the Wolf Creek Generating Station (WCGS).
WC NOC has reviewed the above Reference, which addresses 10 CFR 50.46 reporting information pertaining to the Emergency Core Cooling System (ECCS) Evaluation Model changes that were implemented by Westinghouse for 2018. The review concludes that the effect of changes to, or errors in, the Evaluation Models on the limiting transient peak cladding temperature (PCT) is not significant for 2018. Therefore, changes to the ECCS Evaluation Models are being reported as an annual report.
Attachment I provides an assessment of the specific changes and enhancements to the Westinghouse Evaluation Models for 2018. 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 An Equal Opportunity Employer M/F/HCNET
RA 19-0034 Page 2 of 2 Coolant Attachment II provides PCT rack-up forms for the calculated Large Break Loss-of-ns in effect for the 2018 WCGS Accident (LOCA) and Small Break LOCA PCT margin allocatio the tables, both Cycle 23 Evaluation Models. Since WCNOC has penalties that are carried on dent tables are specific and Cycle 23 independent tables are included. The Cycle 23 indepen Break LOCA labeled as current. The PCT values determined in the Large Break and Small the 10 CFR analyses of record, combined with all of the PCT allocations, remain below re, WCGS is in complia nce with 10 CFR 50.46 50.46(b)(1) regulatory limit of 2200 °F. Therefo requirements and no reanalysis or other action is required.
matter, please This letter contains no commitments. If you have any questions concerning this contact me at (620) 364-4204.
Sincerely, Ron Benham RDB/rlt Attachments: I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Models for Large and Small Break Loss-of-Coolant Accidents (LOCA)
II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Forms cc: S. A. Morris (NRC), w/a B. K. Singal (NRC), w/a N. H. Taylor (NRC), w/a Senior Resident Inspector (NRC), w/a
Attachment I of RA 19-0034 Page 1 of 2 ASSESSMENT OF CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS FOR LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENTS (LOCA)
WOLF CREEK CYCLE 23 FUEL ROD RECONSTITUTION
Background
Within one fuel assembly, a single rod is being reconstituted for Wolf Creek Cycle 23. This item represents a change in plant configuration or associated *set points, distinguished from an evaluation model change in Section 4 of WCAP-13451.
Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The Large Break LOCA ASTRUM and Small Break LOCA NOTRUMP analyses of record were evaluated for one reconstituted rod. The estimated effect is 0°F for Small Break LOCA and 0°F for Large Break LOCA.
U02 FUEL PELLET HEAT CAPACITY
Background
A typographical error was discovered in the implementation of the Uranium Dioxide (U02) fuel pellet heat capacity as described by Equation C-4 of WCAP-8301 [1] for fuel rod heat-up calculations within the Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an over-prediction of heat capacity that increases with fuel temperature. The corrected formulation results in a maximum decrease in heat capacity on the order of approximately 1.2% for existing analyses of record. This 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 The small over-prediction in U02 fuel pellet heat capacity has been evalua*ted to have a negligible effect on existing large and small break LOCA analysis results due to the small magnitude of the change, leading to an estimated peak cladding temperature (PCT) impact of 0°F.
Reference
Attachment I of RA 19-0034 Page 2 of 2 RADIATION HEAT TRANSFER TO LIQUID
Background
It was discovered that under certain *conditions, the radiation heat transfer to liquid could be incorrectly calculated by the thermal-hydraulic codes within the Westinghouse Automated Statistical Treatment of Uncertainty Method (ASTRUM) and FULL SPECTRUM LOCA TM (FSLOCA ') best-estimate LOCA evaluation models. The radiation heat transfer to liquid is generally a small portion of the overall heat transfer. The correction of this error represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Engineering judgement supported by sensitivity calculations exerc1s1ng the heat transfer package showed that correcting this error had minimal impact on LOCA calculations, leading to an estimated PCT impact of 0°F.
FULL SPECTRUM and FSLOCA are trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
VAPOR TEMPERATURE RESETTING
Background
In the WCOBRA/TRAC and WCOBRA/TRAC-TF2 codes, when the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. It was discovered that this vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. The correction of this error represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Engineering judgement supported by sensitivity calculations showed that correcting this error had minimal impact on LOCA transient calculations, leading to an estimated PCT impact of 0°F.
Attachment 11 of RA 19-0034 Page 1 of 5 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORMS LOCA Peak Cladding Temperature (PCT) Summary Plant Name: WOLF CREEK Utility Name: Wolf Creek NOC EM: NOTRUMP AOR
Description:
Appendix K Small Break Summary Sheet Status: Current Reference PCT (°F) Note#
ANALYSIS-OF-RECORD 936 1 Delta PCT Reference Reporting ASSESSMENTS* (o~F) # Note# Year**
- 1. Loose Part Evaluation 45 2 (a) 1990
- AOR + ASSESSMENTS PCT= 981.0 °F
- The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
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-1-90-239, "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.
1 of RA 19-0034 Page 2 of 5 LOCA Peak Cladding Temperature {PCT) Summary Plant Name: WOLF CREEK Utility Name: Wolf Creek NOC EM: NOTRUMP AOR
Description:
Appendix K Small Break Summary Sheet Status: Cycle 23 Reference PCT {°F) Note#
ANALYSIS-OF-RECORD 936 1 Delta PCT Reference
- Reporting 0
ASSESSMENTS* { aF) # Note# Year**
- 1. Loose Part Evaluation 45 2 (a) 1990
- 2. Evaluation of One Reconstituted Fuel Rod 0 3 2018 (Limit 3.02 Violation)
AOR + ASSESSMENTS PCT = 981.0 °F
- The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
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-1-90-239, "Wolf Creek Nuclear Operating Corporation, RCS Loose Part Evaluation," April 1990.
3 LTR-LIS-18-64 Revision 1, "Wolf Creek (SAP) Cycle 23 LOCA Safety Evaluation Input and Potential Issue (Pl) Notification;" April 2018.
NOTES:
(a) This penalty will be carried to track the loose part which has not been recovered.
Attachment II of RA 19-0034 Page 3 of 5 LOCA Peak Cladding Temperature (PCT) Summary Plant Name: WOLF CREEK Utility Name: Wolf Creek NOC EM: ASTRUM (2004)
Description:
Best Estimate Large Break Summary Sheet Status: Current Reference PCT (°F) Note#
ANALYSIS-OF-RECORD 1900 1 Delta PCT Reference Reporting 0
ASSESSMENTS* ( AF) # Note# Year**
- 1. Containment Fan Cooler 2014 0 2 Capacity
- 2. Decay Group Uncertainty
- 2014
-10 3 Factors Errors
- 3. Containment Fan Cooler 2014 0 2 (a)
Capacity AOR + ASSESSMENTS PCT = 1890.0 °F
- The licensee should determine the reportability of these assessments pur~uant to 10 CFR 50.46.
- The "Reporting Year refers to the annual reporting year in which this assessment was included.
REFERENCES 1 WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January 2014.
2 LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.
3 LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.
NOTES:
(a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.
1 of RA 19-0034 Page 4 of 5 LOCA Peak Cladding Temperature {PCT) Summary Plant Name: WOLF CREEK Utility Name: Wolf Creek NOC EM: ASTRUM (2004)
Description:
Best Estimate Large Break Summary Sheet Status: Cycle 23 Reference PCT {°F) Note#
ANALYSIS-OF-RECORD 1900 1 Delta PCT Reference Reporting 0
ASSESSMENTS* ( aF) # Note# Year**
- 1. Containment Fan Cooler 2014 0 2 Capacity
- 2. Decay Group Uncertainty 2014
-10 3 Factors Errors
- 3. Containment Fan Cooler 2014 0 2 (a)
Capacity
- 4. Evaluation of One 2018 Reconstituted Fuel Rod *o 4 (Limit 3.02 Violation)
AOR + ASSESSMENTS PCT = 1890.0 °F
- The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
REFERENCES 1 WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January* 2014.
2 LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.
3 LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.
4 LTR-LIS-18-64, Revision 1, "Wolf Creek (SAP) Cycle 23 LOCA Safety Evaluation Input and Potential lss*ue (Pl) Notification," April 2018.
Attachment II of RA 19-0034 Page 5 of 5 NOTES:
(a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.