ML18142A072

Provides Info Re Reload Design & non-LOCA Safety Analysis Methodology Based on Westinghouse Computer Codes,Per NRC 840619 Request.Info Re Training Encl

ML18142A072
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	10/09/1984
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Eisenhut D Office of Nuclear Reactor Regulation
References
553, NUDOCS 8410150302
Download: ML18142A072 (5)

Text

e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 W. L. STEWART VICE PRESIDENT NUCLEAR OPERATIONS October 9, 1984 Mr. Harold R. Denton, Director Serial No.: 553 Office of Nuclear Reactor Regulation E&C/DD:klh:2037C Attn: Mr. D. G. Eisenhut, Director Docket Nos: 50-280, 50-281 Division of Licensing 50-338, 50-339 U.S. Nuclear Regulatory Commission License Nos: DPR-32, DPR-37 Washington, D.C. 20555 NPF-4, NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 USE OF WESTINGHOUSE COMPUTER CODES FOR RELOAD DESIGN AND NON-LOCA SAFETY ANALYSIS BY VEPCO On May 16 and 17, 1984 NRC Representatives accompanied by technical consultants from Brookhaven National Laboratory (BNL) conducted an audit at the Vepco offices in Richmond, Virginia. The purpose of the audit was to assess Vepco s capability to perform reload design and non-LOCA safety analyses using 1

Westinghouse computer codes. Because of the use of Westinghouse computer codes by Vepco is only on an interim basis, and the codes are being used for near term reload analyses, it was decided that an evaluation could be performed with an audit rather than through the process of preparation and review of topical reports.

On June 19, 1984, Mr. James R. Miller of Operating Reactors Branch #3 of the Division of Licensing formally transmitted by letter the results of the audit. The evaluation (provided as Enclosure 1 to Mr. Miller's letter) found that Vepco s use of the Westinghouse computer codes for reload design and 1

non-LOCA safety analysis was acceptable on an interim basis. Also, the evaluation cited that it was the NRC s understanding that Westinghouse computer 1

codes will be phased out as approval of Vepco computer code development progresses.

Finally, as stated in the evaluation, the NRC requested two letters from Vepco, one of which would address Quality Assurance and the other would address the Vepco use of Westinghouse computer codes for reload design and safety analysis. This letter addresses Vepco s reload design and non-LOCA safety 1

analysis methodology based on Westinghouse computer codes, the calculations performed and the results obtained, the codes used, and the check reviews performed.

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e e VIROINIA. ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton, Director The basis of Vepco s reload design and non-LOCA safety analysis procedures 1

is WCAP-9272 (Westinghouse Reload Safety Evaluation Methodology dated March 1978). All safety related key nuclear parameters are examined on a cycle to cycle and design specific basis according to the methodology presented in WCAP-9272. The non-LOCA safety analysis is based on the reference analysis approach as described in WCAP-9272. This approach determines if a core configuration is bounded by existing safety analyses in order to confirm that applicable safety criteria are satisfied. When a key nuclear parameter is not bounded, the postulated accidents affected by that parameter are either re-evaluated or reanalyzed. If the accident is reanalyzed, standard non-LOCA safety analysis methods are followed and analytical assumptions which are consistent with the licensing bases in our Updated Final Safety Analysis Report (UFSAR) are employed.

Vepco s understanding and use of Westinghouse computer codes as described 1

in WCAP-9272 has been further augmented by extensive training in the reload design and non-LOCA safety analysis areas as shown in Attachment 1. Further training and experience on the use of Westinghouse computer codes was gained by completing reload designs for Surry 2 Cycle 6, North Anna 2 Cycle 2, North Anna 1 Cycle 4, and Surry 1 Cycle 7 in parallel with Westinghouse. In addition, several non-LOCA safety analyses were performed and compared to Westinghouse North Anna core uprating feasibility study analyses.

The nuclear calculations performed by Vepco using Westinghouse computer codes are described in WCAP-9272. These are identified in Attachment 2.

In addition, an evaluation is performed to confirm that the core thermal limits remain conservative and that no significant variations in thermal margins result from the core reload. This includes both steady state and transient evaluations as required.

The primary Westinghouse computer codes used by Vepco in determining key nuclear parameters are the same as those used currently by Westinghouse. They are TORTISE (2-D diffusion), THURTLE (3-D diffusion), PALADON (2-D and 3-D nodal), APOLLO (1-D diffusion), and ARK (cross sections) which are similar to the TURTLE, FLARE, PANDA and LEOPARD/CINDER codes described in WCAP-9272.

Non-LOCA safety analysis codes available for use by Vepco include LOFTRAN, FACTRAN, TWINKLE and THINC. .

After the key nuclear parameters and any required safety analyses are determined using the Westinghouse computer codes, checks are made to ensure that all of the calculations have been done properly. Computer input and hand calculations made to derive key nuclear parameters from calculated results are independently reviewed by cognizant engineers to ensure correctness regarding input and calculational procedures. The engineers reviewing the input to the calculations and the results have not been involved in the specific cycle design process but are knowledgeable in the appropriate design or safety analysis area. All calculations are documented in calculational notes and/or technical reports and are signed by the preparer and reviewers once the quality and accuracy of the calculation has been confirmed.

e e VIRGINIA. ELECTRIC AND POWER COJ,{PANY TO Mr. Harold R. Denton, Director If you have any further questions on the use of Westinghouse computer codes by Vepco, please contact us.

Very truly yours,

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W. L. Stewart DD:klh:2037C Attachments cc: Mr. Daniel Fieno Core Performance Branch Office of Nuclear Reactor Regulation Mr. M. W. Branch NRC Resident Inspector North Anna Power Station Mr. D. J. Burke NRC Resident Inspector Surry Power Station Mr. James R. Miller, Chief Operating Reactors Branch No. 3 Division of Licensing Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing Mr. J. P. 0 Reilly 1

Regional Administrator Region II

e ATTACHMENT 1 TRAIN ING WESTINGHOUSE DESIGN AND SAFETY ANALYSIS CODES AND METHODOLOGY Course Course Number of Total Description Length Presentations Training Introduction to the W 3 Days 2 39 Man-Days Computer System Basic PWR Core Physics 5 Days 1 35 Man-Days

, and Therma 1 Hydraulics Methodology and Computer 5 Days 3 80 Man-Days

• Models for WDesign Codes Westi~ghouse APOLLO and 5 Days 2 60 Man-Days MIXIN Design Codes Nuclear Design Development 5 Days 3 115 Man-Days of the Reload Safety Analysis Checklist Introduction to Non-LOCA 5 Days 3 80 Man-Days Safety Analysis Rod Ejection, Main Steam- 5 Days 2 40 Man-Days line Break, Dropped Rod Analysis Westinghouse Thermal 5 Days 1 20 Man-Days Hydraulic Methods Westinghouse Large Break 5 Days 2 30 Man-Days LOCA Codes (Theory)

Westinghouse Large Break 20 Days 3 60 Man-Days LOCA Codes (On-The-Job-Training) 559 Man-Days 73-DD-2037C-5

i e e ATTACHMENT 2 NUCLEAR CALCULATIONS PERFORMED BY VEPCO USING WESTINGHOUSE COMPUTER CODES Core Reactivity Parameters and Coefficients moderator temperature (density) coefficients, doppler, temperature and power coefficients, boron worth, effective delayed neutron fraction, prompt neutron lifetime.

Control Rod Worth Parameters - Rod insertion limits, shutdown margin trip reactivity, differential rod worth.

Key Parameters for Specific Events limiting core power distribution during normal operation, base load and load follow operation; rod control equipment malfunctions; boron dilution during power operation, at hot shutdown, and during refueling; single rod withdrawal during full power operation; statically misaligned rod during power operations; dropped rod and bank during full power operation; rod ejection; steamline break 73-DD-2037C-6