L-19-075, Submittal of 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models

Submittal of 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models

ML19106A007
Person / Time
Site:	Beaver Valley
Issue date:	04/15/2019
From:	Bologna R FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-19-075
Download: ML19106A007 (5)

Text

FENOC Beaver Valley Power Station

~ P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 April 15, 2019 L-19-075 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington , DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334 , License No. DPR-66 Docket No. 50-412, License No. NPF-73 2018 Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models In accordance with Title 10 of the Code of Federal Regulations, Part 50, Section 50.46(a)(3)(ii), FirstEnergy Nuclear Operating Company (FENOC) hereby submits the 2018 annual report of changes to or errors in emergency core cooling system evaluation models, or in the application of the models, for the Beaver Valley Power Station , Unit Nos. 1 (BVPS-1) and 2 (BVPS-2).

The attachments provide a summary list and description of each change to or error in the acceptable evaluation models that affects the peak fuel cladding temperature (PCT) calculation for various loss-of-coolant accidents, as well as the estimated PCT effects of the change or error.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required , please contact Mr. Phil H. Lashley, Acting Manager - Nuclear Licensing and Regulatory Affairs, at (330) 315-6808.

Sincerely,

~;M --

Richard D. Bologna

Beaver Valley Power Station , Unit Nos. 1 and 2 L-19-075 Page 2 Attachments:

1. Summary of 2018 Peak Fuel Cladding Temperature (PCT) Effects for Beaver Valley Power Station (BVPS) Loss-of-Coolant Accident (LOCA) Transients

2. Descriptions of 2018 Emergency Core Cooling System (ECCS) Evaluation Model Changes or Errors cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment 1 L-19-075 Summary of 2018 Peak Fuel Cladding Temperature (PCT) Effects for Beaver Valley Power Station (BVPS) Loss-of-Coolant Accident (LOCA) Transients Page 1 of 1 Beaver Valley Power Station, Unit 1 Large Break Small Break LOCA LOCA PCT or PCT PCT or PCT ChanQe ChanQe Licensing Basis PCT at BEGINNING of 2018 1840°F 1895°F 2018 Activity EM ChanQes None NIA NIA EM Errors U02 Fuel Pellet Heat Capacity (refer to NIA 0°F page 1 of Attachment 2)

Vapor Temperature Resetting (refer to 0°F NIA page 2 of Attachment 2)

Licensing Basis PCT at END of 2018 1840°F 1895°F Beaver Valley Power Station, Unit 2 Large Break Small Break LOCA LOCA PCT or PCT PCT or PCT Chanoe Change Licensing Basis PCT at BEGINNING of 2018 1839°F 1917°F 2018 Activity EM Chanoes None NIA NIA EM Errors U02 Fuel Pellet Heat Capacity (refer to NIA 0°F page 1 of Attachment 2)

Vapor Temperature Resetting (refer to 0°F NIA page 2 of Attachment 2)

Licensing Basis PCT at END of 2018 1839°F 1917°F

Attachment 2 L-19-075 Descriptions of 2018 Emergency Core Cooling System (ECCS)

Evaluation Model Changes or Errors Page 1 of 2 Uranium Dioxide (U02) FUEL PELLET HEAT CAPACITY

Background

A typographical error was discovered in the implementation of the UO2 fuel pellet heat capacity as described by Equation C-4 of WCAP-8301, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," June 1974, for fuel rod heat-up calculations within Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an overprediction 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 percent for existing analyses of record. This represents a non-discretionary change in accordance with Section 4.1.2 of WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

Affected Evaluation Model(s)

1. 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP (Applicable to BVPS-1 and BVPS-2)

Estimated Effect The small over-prediction in UO2 fuel pellet heat capacity has been evaluated 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 PCT impact of O degrees Fahrenheit (°F) for both BVPS-1 and BVPS-2.

L-19-075 Page 2 of 2 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)

1. 1996 Westinghouse Best-Estimate Large Break LOCA Evaluation Model (Applicable to BVPS-2)

2. 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM (Applicable to BVPS-1)

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 for both BVPS-1 and BVPS-2.