L-08-340, Transmittal of 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models

Transmittal of 10 CFR 50.46 Report of Changes or Errors in ECCS Evaluation Models

ML083390106
Person / Time
Site:	Beaver Valley
Issue date:	11/26/2008
From:	Sena P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-340, TAC MC3394, TAC MC3395, TAC MC4645, TAC MC4646, TAC MC4648
Download: ML083390106 (8)

Text

FENOC ,n

• Beaver Valley PowerStation P.O. Box 4 FirstEnergyNuclear OperatingCompany Shippingport, PA 15077 PeterP. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 November 26, 2008 L-08-340 10 CFR 50.46(a)(3)(ii)

ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BY-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 10 CFR 50.46 Report of Chanqes or Errors in ECCS Evaluation Models Pursuant to 10 CFR 50.46(a)(3)(ii), FirstEnergy Nuclear Operating Company (FENOC) provides the attached report as annual notification of changes or errors in emergency core cooling system (ECCS) evaluation models for Beaver Valley Power Station (BVPS)

Unit Nos. 1 and 2. Current information for both large and small break transients is provided to satisfy reporting requirements. The following attachments provide information as requested by 10 CFR 50.46:

Attachment I Provides a listing of each change or error in an acceptable evaluation model that affects the peak cladding temperature (PCT) calculation for particular transients. It quantifies the effects of changes that have occurred since the previous report (December 4, 2007) for the specified transients and provides an index into Attachment 2.

Attachment 2 Provides a description for each model change or error.

The PCT effects, listed in Attachment 1, result in PCTs for the large and small break loss of coolant accident (LOCA) transients as follows:

BVPS-1 Large Break LOCA - 2014OF BVPS-1 Small Break LOCA - 1895 0F BVPS-2 Large Break LOCA - 20171F BVPS-2 Small Break LOCA - 1917°F Changes or errors reflected in the PCT values above include those previously reviewed and approved by the NRC via license amendments associated with extended power uprate (TAC Nos. MC4645 and MC4646), containment conversion (TAC Nos. MC3394 1i1OD1

Beaver Valley Power Station, Unit Nos. 1 and 2 L-08-340 Page 2 and MC3395), and best-estimate loss of coolant accident methodologies (TAC Nos.

MC4647 and MC4648) for BVPS Unit Nos. 1 and 2 as well as those described in previous 10 CFR 50.46 reports provided through December 4, 2007.

FENOC previously committed to performing and submitting a re-analysis of the large break LOCA for BVPS Unit No. 1 within two fuel cycles following implementation of containment conversion (spring 2009) because analysis input changes resulted in PCT impacts of greater than 50 degrees Fahrenheit. This schedule has not changed.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -

Fleet Licensing, at 330-761-6071.

Sincerely, Peter P. Sena III Attachments:

1. Summary of PCT Effects for BVPS LOCA Transients

2. Descriptions of Model Changes or Errors cc: NRC Region I Administrator NRC Senior Resident Inspector NRR Project Manager Director BRP/DEP Nuclear Safety Specialist BRP/DEP

L-08-340 ATTACHMENT 1

SUMMARY

OF PCT EFFECTS FOR BVPS LOCA TRANSIENTS Pcr ATTACHMENT 2 DESCRIPTION EFFECT ('F) PAGE BVPS-1 LARGE BREAK LOCA COUNTER-CURRENT FLOW LIMIT (CCFL) GLOBAL VOLUME ERROR 0 1 HOTSPOT BURST TEMPERATURE LOGIC ERRORS 0 2 BVPS-1 SMALL BREAK LOCA PUMP WEIR RESISTANCE MODELING 0 3 ERRORS IN REACTOR VESSEL LOWER PLENUM SURFACE AREA CALCULATIONS 0 4 BVPS-2 LARGE BREAK LOCA COUNTER-CURRENT FLOW LIMIT (CCFL) GLOBAL VOLUME ERROR 0 1 HOTSPOT BURST TEMPERATURE LOGIC ERRORS 0 2 BVPS-2 SMALL BREAK LOCA PUMP WEIR RESISTANCE MODELING 0 3 ERRORS IN REACTOR VESSEL LOWER PLENUM SURFACE AREA CALCULATIONS 0 4

L-08-340 ATTACHMENT 2 DESCRIPTIONS OF MODEL CHANGES OR ERRORS of L-08-340 Page 1 of 4 COUNTER-CURRENT FLOW LIMIT (CCFL) GLOBAL VOLUME ERROR Backqround An error was identified during the course of a Best Estimate Large Break LOCA analysis in which the volume between the core barrel and the baffle plates in the CCFL region above the active fuel length was modeled incorrectly. The corrected values have been evaluated for impact on the current licensing basis analysis results.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The CCFL global volume modeling error has been generically evaluated to have a negligible impact on PCT for affected analyses and a penalty of 0°F is assigned.

of L-08-340 Page 2 of 4 HOTSPOT BURST TEMPERATURE LOGIC ERRORS

Background

The HOTSPOT code has been updated to incorporate the following corrections to the burst temperature logic: (1) change the rod internal pressure used to calculate the cladding engineering hoop stress from the value in the previous time step to the value in the current time step; (2) revise the average cladding heatup rate calculation to reset selected variables to zero at the beginning of each trial and use the instantaneous heat-up rate when fewer than five values are available; and, (3) reflect the assumed saturation of ramp rate effects above 28°C/s for Zircaloy-4 cladding from Equation 7-66 of Reference 1.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Sample calculations for each change showed no effect on peak cladding temperature, leading to an estimated impact of 0°F for 10 CFR 50.46 reporting purposes.

Reference 1 -WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2-5 (Revision 1),

"Code Qualification Document for Best Estimate LOCA Analysis," S. M. Bajorek et al.,

March 1998.

b Attachment 2 of L-08-340 Page 3 of 4 PUMP WEIR RESISTANCE MODELING

Background

Review of the reactor coolant pump data collections identified instances of either including a weir resistance for a design without a weir or double-counting the weir resistance for a design with a weir. The corrected resistances have been evaluated for impact on existing analysis results and will be incorporated into the plant-specific input databases on a forward-fit basis.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Resolving the identified discrepancies has been evaluated as having a negligible effect on existing results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

of L-08-340 Page 4 of 4 ERRORS IN REACTOR VESSEL LOWER PLENUM SURFACE AREA CALCULATIONS

Background

Two errors were discovered in the calculations of reactor vessel lower plenum surface area. The corrected values have been evaluated for impact on current licensing-basis analysis results and will be incorporated on a forward-fit basis.

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 vessel lower plenum surface area are relatively minor and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 0F for 10 CFR 50.46 reporting purposes.