RNP-RA/09-0035, Report of Changes to or Errors Discovered in an Acceptable Loss-of-coolant Accident Evaluation Model Application for the Emergency Core Cooling System

Report of Changes to or Errors Discovered in an Acceptable Loss-of-coolant Accident Evaluation Model Application for the Emergency Core Cooling System

ML091110317
Person / Time
Site:	Robinson
Issue date:	04/16/2009
From:	Castell C Carolina Power & Light Co, Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/09-0035
Download: ML091110317 (4)

Text

P10 SProgress Energy CFR 50.46(a)(3)(ii)

Serial: RNP-RA/09-0035 APR 1 12009 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 REPORT OF CHANGES TO OR ERRORS DISCOVERED IN AN ACCEPTABLE LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL APPLICATION FOR THE EMERGENCY CORE COOLING SYSTEM Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.46, Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), is submitting the attached report of changes to and errors discovered in an acceptable Loss-of-Coolant Accident (LOCA) evaluation model (EM) for the Emergency Core Cooling System at the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The applicable LOCA EMs are referenced in the HBRSEP, Unit No. 2, Core Operating Limits Report. Changes to or errors discovered in EMs and EM applications were previously reported to the Nuclear Regulatory Commission by letter dated December 2, 2008.

This submittal satisfies the notification of a significant change [greater than 50'F change in calculated peak cladding temperature (PCT)] for the Small Break LOCA (SBLOCA), as required by 10 CFR 50.46(a)(3)(ii). Although the change in the Large Break LOCA (LBLOCA) is less than 50'F, the impact on the LBLOCA is reported for completeness, as the changed condition is the same as that for the SBLOCA. The change condition and PCT impact are provided in Attachment I.

The latest PCT estimates for the LBLOCA and SBLOCA are included in Attachment II.

There is no reanalysis planned because the changes result in a net improvement in margin to PCT limits. The cumulative impact of changes will continue to be tracked.

If you have any questions concerning this matter, please contact me at (843) 857-1626.

Sincerely, Curt Castell Supervisor - Licensing/Regulatory Programs Progress Energy Carolinas, Inc.

Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

United States Nuclear Regulatory Commission Serial: RNP-RA/09-0035 Page 2 of 2 RAC/rac Attachments:

I. Report of Changes/Errors in Loss-of-Coolant Accident Evaluation Models for the Emergency Core Cooling System Ii. Peak Cladding Temperature Estimates c: L. A. Reyes, NRC, Region II M. G. Vaaler, NRC, NRR NRC Resident Inspector, HBRSEP

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/09-0035 Page 1 of 1 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REPORT OF CHANGES/ERRORS IN LOSS-OF-COOLANT ACCIDENT EVALUATION MODELS FOR THE EMERGENCY CORE COOLING SYSTEM This report provides an estimate of the effect on peak cladding temperature (PCT) of changes and error corrections in the Loss-of-Coolant Accident (LOCA) evaluation models (EMs) and EM applications for the Emergency Core Cooling System (ECCS) at the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, covering the period of November 11, 2008 through March 24, 2009.

Large Break Loss-of-Coolant Accident (LBLOCA) Evaluation Model CHANGED CONDITION PCT IMPACT ('F)

An incorrect figure for predicting radiation heat transfer was incorporated into RELAP approximately 30 years ago. This error was incorporated into S-RELAP5 and impacts its radiation to fluid -32 heat transfer model. The S-RELAP5 code is used to analyze a Large Break LOCA.

Cumulative Impact -32 Small Break Loss-of-Coolant Accident (SBLOCA) Evaluation Model CHANGED CONDITION PCT IMPACT ('F)

An incorrect figure for predicting radiation heat transfer was -64 incorporated into RELAP approximately 30 years ago. This error was incorporated into ANF-RELAP and impacts its radiation to fluid heat transfer model. The ANF-RELAP code is used to analyze small break LOCA.

Cumulative Impact -64

United States Nuclear Regulatory Commission I to Serial: RNP-RA/09-0035 Page 1 of 1 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PEAK CLADDING TEMPERATURE ESTIMATES The current peak cladding temperature (PCT) estimates associated with Loss-of-Coolant Accident (LOCA) Emergency Core Cooling System (ECCS) evaluation models are listed below. These estimates include the cumulative effects of significant and non-significant error corrections and evaluation model changes through March 24, 2009.

Event PCT (OF)

Large Break LOCA, ECCS Injection Mode 1853 Small Break LOCA, ECCS Injection Mode 1607