ML18153C309

Provides Outline of Plan to Meet 10CFR50 App G Requirements, for Low Upper Shelf Energy Matls,Per NRC 900521 Request

ML18153C309
Person / Time
Site:	Surry
Issue date:	07/30/1990
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-335, NUDOCS 9008030066
Download: ML18153C309 (3)

Text

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,* e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 July 30, 1990 United States Nuclear Regulatory Commission Serial No.90-335 Attention: Document Control Desk NES/JRH/cdk/vlh Washington, D. C. 20555 Docket No. 50-280 License No. DPR-32 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 1 RESPONSE TO REQUEST FOR PLAN TO MEET REQUIREMENTS OF 10 CFR 50 APPENDIX G FOR LOW UPPER SHELF ENERGY MATERIALS By letter dated December 1, 1989, "Response to Request for Additional Information; Upper Shelf Energy of Reactor Vessel Materials," Virginia Electric and Power Company (the Company) reported to NRC staff that the upper shelf energy (USE) of the lead material in the beltline region is conservatively predicted to reach 50 ft-lbs at the 1/4-T location in 1998. This letter also provided a description of the Company's actions in response to the low upper shelf toughness issue. NRC staff subsequently requested by letter dated May 21, 1990, "Surry Unit 1 - Low Upper Shelf Material in Reactor Vessel Beltline," that the Company provide a plan for submitting the fracture mechanics analysis and performing the volumetric examination in accordance with the requirements of 10 CFR 50 Appendix G. The purpose of this letter is to provide you with an outline of our plan to meet the Appendix G requirements.

For those reactor vessel beltline materials which no longer meet the 50 ft-lb USE criterion, the following requirements must be satisfied:

1. A volumetric inspection of 100% of the beltline materials that do not satisfy the requirements of Section V.B of Appendix G must be performed and any flaws must be characterized according to Section XI of the ASME Code and as otherwise specified by the Director, Office of Nuclear Regulation.

2. Additional evidence of the fracture toughness of the beltline materials after exposure to neutron irradiation must be obtained from results of supplemental fracture toughness tests.

3. An analysis must be performed that conservatively demonstrates, making appropriate allowances for uncertainties, the existence of equivalent margins of safety for continued operation.

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e During the Surry Unit 1 Cycle 12/13 refueling outage currently scheduled for fall of 1993, the Company is planning to conduct a 100% volumetric inspection of the reactor vessel beltline materials that are conservatively predicted to drop below the 50 ft-lb USE criterion between the present time and the end of the Unit 1 vessel's currently licensed life. Th.ese materials will be identified by the calculational method prescribed in Regulatory Guide 1.99, Revision 2. The reactor vessel examination will be conducted in accordance with the guidelines of the ASME Boiler and Pressure Vessel Code.

Additional evidence of the fracture toughness of the Surry Unit 1 reactor vessel beltline materials will be provided through our participation in the activities of the Reactor Vessel Working Group of the Babcock and Wilcox Owner's Group (B&WOG). The B&WOG has obtained NRC approval of fracture toughness testing procedures (Reference 1), a fracture mechanics analysis procedure (Reference 2), and an integrated reactor vessel surveillance program (Reference 3). Three years prior to exceeding the 50 ft-lb USE criterion, as conservatively predicted by Regulatory Guide 1.99, Revision 2, a plant specific fracture toughness analysis for Surry Unit 1 will be provided. This analysis will be supported by experimental and analytical methods which have been reviewed and approved by NRC staff, and will include data from the integrated reactor vessel surveillance program which incorporates the materials and conditions found in the Surry Unit 1 reactor vessel. The analysis is expected to conservatively demonstrate, with appropriate allowances for uncertainties, the existence of equivalent margins of safety for continued operation of the Surry Unit 1 vessel.

The Company is continuing work on a flux reduction study for Surry Unit 1. The intent of the program is to reduce the neutron flux at the critical locations (i.e., certain welds) in order to maintain the neutron fluence below levels that would result in exceeding material integrity safety limits. Design calculations indicate that the desired level of flux reduction can be achieved using part length absorber in the control rod guide tubes of the peripheral assemblies located near the critical locations. We are presently planning to insert the part length absorbers in the Surry 1, Cycle 12 or 13 reload core.

Further details of the programs being instituted to meet the requirements of Appendix G, Paragraph V.C will be submitted in accordance with the requirements of Appendix G, Paragraph V.E.

Should you have further comments or questions, please contact us.

Very truly yours, Ul.l\.~

W. L. Stewart Senior Vice President - Nuclear Attachments

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References:

(1) J. D. Aadland, et al.: "Babcock and Wilcox J-R Procedure for Compact Fracture Toughness Specimens," B&W-1808, Babcock and Wilcox, Lynchburg, Virginia (December, 1984).

(2) "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," B&W-10046, Revision 2, Babcock and Wilcox, Lynchburg, Virginia (December, 1984).

(3) A. L. Lowe, Jr., et al.: "Integrated Reactor Vessel Material Surveillance Program," B&W-1543 A, Revision 2, Babcock and Wilcox, Lynchburg, Virginia (May, 1985).

cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Mr. Donald Howell c/o Babcock & Wilcox P. 0. Box 10935 Lynchburg, Virginia 24506-0935