NL-05-058, Reanalysis of Large Break Loss of Coolant Accident Using Astrum

Reanalysis of Large Break Loss of Coolant Accident Using Astrum

ML051230311
Person / Time
Site:	Indian Point
Issue date:	04/22/2005
From:	Dacimo F Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-05-058
Download: ML051230311 (4)

Text

Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB En tergyP0. Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration April 22, 2005 Re: Indian Point Unit No. 2 Docket No. 50-247 NL-05-058 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Reanalysis of Large Break Loss of Coolant Accident Using ASTRUM

References:

1. Entergy letter to NRC (NL-04-081), "Proposed Schedule for Reanalysis of Large Break Loss of Coolant Accident", dated July 2, 2004.

2. NRC letter to Westinghouse Electric Company, "Final Safety Evaluation for WCAP-16009-P, Revision 0, 'Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)", dated November 5, 2004.

Dear Sir:

Entergy Nuclear Operations, Inc (Entergy) has completed a reanalysis of the Large Break Loss of Coolant Accident (LBLOCA) for Indian Point Unit 2 (IP2), as committed in Reference 1.

The reanalysis applies the Westinghouse ASTRUM methodology recently approved by NRC (Reference 2) and is based on a full core of upgraded fuel at an operating power level of 3216 MWth. Analysis results are summarized in Attachment 1. The current operating cycle for IP2 is the first intermediate fuel cycle for the transition to the upgraded fuel assembly design.

Therefore, as committed in Reference 1, a transitional assessment was performed and the resulting transition core penalty is reported in Attachment 1.

Entergy plans to submit a license amendment request to NRC by October 2005 regarding the adoption of this methodology for IP2. Reporting in accordance with 10 CFR 50.46 will continue to be based on the current licensing basis analysis methodology until the proposed new methodology is approved for implementation at IP2.

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NL-05-058 Docket No. 50-247 Page 2 of 2 There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Kevin Kingsley at 914-734-6695.

Sincerely, re R>. Dacimo Site Vice President Indian Point Energy Center cc: Mr. Patrick D. Milano, Senior Project Manager Project Directorate 1, Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission Mail Stop 0 8 C2 Washington, DC 20555 Mr. Samuel J. Collins Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspectors Office Indian Point Unit 2 U.S. Nuclear Regulatory Commission P.O. Box 59 Buchanan, NY 10511 Mr. Peter R. Smith, President New York State Energy, Research and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223

ATTACHMENT I TO NL-05-058 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO THE INDIAN POINT UNIT 2 NUCLEAR PLANT The original application of the Westinghouse Best Estimate Methodology to the Indian Point Unit 2 Nuclear Plant (IP2) was approved by the NRC in 1997 (Reference 1). The original application employed the NRC approved 1996 Evaluation Model (Reference 2).

Westinghouse recently underwent a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is still based on the CQD methodology (Reference 2) and follows the steps in the CSAU methodology.

However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case.

The ASTRUM evaluation model is documented in WCAP-1 6009-P-A (Reference 3).

The WCOBRAJTRAC model for IP2 was originally developed at 3216 MWth. The WCOBRA/TRAC noding that was developed at that time remains unchanged for the BELOCA ASTRUM analysis.

The ASTRUM BELOCA analysis was performed for a full core of upgraded fuel. The analysis was performed in compliance with all the NRC conditions and limitations as identified in WCAP-1 6009-P-A (Reference 3).

Table 1 summarizes the results of the ASTRUM BELOCA analysis. Based on the results, it is concluded that the Indian Point Unit 2 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

Table 1: Indian Point Unit 2 ASTRUM Best Estimate Large Break LOCA Results 10 CFR 50.46 Parameter BELOCA Result Criteria 95/95 PCT (uF) 1962 < 2200 95 / 95 Local Max Oxidation (%) 2.39 < 17 95 / 95 Core Wide Oxidation (%) 0.35 <1 Coolable Geometry Core remains coolable Core remains coolable Long Term Cooling Core remains cool in long Core remains cool in long Iterm term

References:

1. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

188 to Facility Operating License No. DPR-26 Consolidated Edison Company of New York, Inc, Indian Point Unit 2, Docket No. 50-247, March 1997.

2. Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis,"WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).

3. Nissley, M. E., et.al., 2005, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A.

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ATTACHMENT I TO NL-05-058 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Plant Name: Indian Point Unit 2 Future Utility Name: Entergy Nuclear Northeast Revision Date: 4/4/05 Analvsis Information EMl: WCOBRAfTRAC Analysis Date: 2/15/05 Limiting Break Size: Guillotine FQ: 2.5 Fdll: 1.7 Fuel: 15x15 Upgraded SGTP (%): 10 Notes:

Clad Temp (0F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1962 1 MARGIN ALLOCATIONS (Delta PCT)

A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS 1 .None 0 B. PLANNED PLANT CHANGE EVALUATIONS I . Bent Fuel Assembly Alignment Pins 5 1 C. 2004 PERMANENT ECCS MODEL ASSESSMENTS I .None 0 D. TEMPORARY ECCS MODEL ISSUES*

I . Transition Core Penalty 26 1 (a) 2 . Cycle 17 Typical Cycle Average Bumup 0 2 (b)

E. OTHER I . None 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1993

• It is recommended that these temporary PCT allocations which address current LOCA model issues not be considered with respect to 10 CFR 50.46 reporting requirements.

References:

1. WCAP-16405-P (DRAFT), "Best Estimate Analysis of the Large Break Loss ofCoolant Accident for Indian Point Unit 2 Nuclear Plant Using the ASTRUM Methodology." March 2005.

2. IPP-05-17. 10 CFR 50.46 Annual Notification and Reporting for 2004," April2005.

Notes:

(a) Transition core penalty will be in effect during the transition cycles only. Once a full core of 15x15 upgraded fuel is loaded, then the penalty would be no longer applicable.

(b) Typical Cycle Average Bumup assessment will be in ceffect during the cycle 17 operation only. Once the cycle 18 begins, then the assessment would be no longer applicable.

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