ML020510009

Technical Specification Basis Change, Alternate Source Term Implementation

ML020510009
Person / Time
Site:	Surry
Issue date:	02/08/2002
From:	Grecheck E Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
01-037D
Download: ML020510009 (6)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 8, 2002 U. S. Nuclear Regulatory Commission Serial No. 01-037D Attention: Document Control Desk SPS-LIC/CGL R1 Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 TECHNICAL SPECIFICATION BASIS CHANGE ALTERNATE SOURCE TERM IMPLEMENTATION In a letter dated April 11, 2000 (Serial No.00-123), Virginia Electric and Power Company (Dominion) submitted a license amendment request for implementation of the Alternate Source Term (AST) as the plant design and licensing bases for Surry Power Station Units 1 and 2. In a subsequent letter dated July 31, 2001 (Serial No. 01 -037A),

we provided a revised AST analysis report reflecting changes in certain analytical assumptions and results that were incorporated in response to review questions received from the NRC staff. This subsequent letter also indicated that a revision to the proposed Basis section of Technical Specification (TS) 3.10 was required based on the additional analysis work and that this Basis revision would be provided in a later submittal.

The purpose of this letter is to provide the TS 3.10 Basis revision. This revision reflects the following two changes in the TS 3.10 Basis discussion of the fuel handling accident:

1) reference to Regulatory Guide 1.183 (versus the initial reference to the draft guidance of DG 1081) and 2) a revision addressing the revised gap fraction assumptions used in the reanalysis. A UFSAR change request is also being completed consistent with this TS Basis revision and will be included in the next scheduled UFSAR update in accordance with 10CFR50.71 (e).

The TS 3.10 Basis change has been reviewed and approved by the Station Nuclear Safety and Operating Committee.

The marked-up TS Basis page and the proposed TS Basis page are provided in Attachments 1 and 2, respectively. Note that the marked-up page in Attachment 1 is a revision of the marked up page sent in our April 11, 2000 letter. The double revision bar on the marked up page in Attachment 1 identifies the revisions due to the reanalysis submitted by our July 31, 2001 letter.

Should you have any questions regarding the TS Basis change, please contact us.

Very truly yours, E. S. Grecheck Vice President - Nuclear Support Services Attachments:

1. Marked-up TS 3.10 Basis Page

2. Proposed TS 3.10 Basis Page Commitments made in this letter: None.

cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23 T85 Atlanta, Georgia 30303 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station Commissioner Department of Radiological Health Room 240 1500 East Main Street Richmond, VA 23218

ATTACHMENT 1 (to Serial No. 01-037D)

Marked-up TS 3.10 Basis Page Surry Power Station Units 1 and 2 Dominion

TS 3.10-7 XX-XX-XXO8 03 95 Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests will be performed prior to initial criticality.

The fuel handling accident has been analyzed based on the methodology outlined in Draft Regulatory Guide DG 10- -1.2-5 1.183. The analysis assumes 100% release of the gap activity from the highes pew assembly with maximum gap activity i-srelease after a 100-hour decay period following operation at 2605 MWt.

Detailed procedures and checks insure that fuel assemblies are loaded in the proper locations in the core. As an additional check, the movable incore detector system will be used to verify proper power distribution. This system is capable of revealing any assembly enrichment error or loading error which could cause power shapes to be peaked in excess of design value.

References UFSAR Section 5.2 Containment Isolation UFSAR Section 6.3 Consequence Limiting Safeguards UFSAR Section 9.12 Fuel Handling System UFSAR Section 11.3 Radiation Protection UFSAR Section 13.3 Table 13.3-1 UFSAR Section 14.4.1 Fuel Handling Accidents FSAR Supplement: Volume I: Question 3.2 Amendment Nos. XXX20 and XXX20-

ATTACHMENT 2 (to Serial No. 01-037D)

Proposed TS 3.10 Basis Page Surry Power Station Units 1 and 2 Dominion

TS 3.10-7 Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests will be performed prior to initial criticality.

The fuel handling accident has been analyzed based on the methodology outlined in Regulatory Guide 1.183. The analysis assumes 100% release of the gap activity from the assembly with maximum gap activity after a 100-hour decay period following operation at 2605 MWt.

Detailed procedures and checks insure that fuel assemblies are loaded in the proper locations in the core. As an additional check, the movable incore detector system will be used to verify proper power distribution. This system is capable of revealing any assembly enrichment error or loading error which could cause power shapes to be peaked in excess of design value.

References UFSAR Section 5.2 Containment Isolation UFSAR Section 6.3 Consequence Limiting Safeguards UFSAR Section 9.12 Fuel Handling System UFSAR Section 11.3 Radiation Protection UFSAR Section 13.3 Table 13.3-1 UFSAR Section 14.4.1 Fuel Handling Accidents FSAR Supplement: Volume I: Question 3.2 Amendment Nos.