ML18153C413

Forwards Info Necessary to Complete Assessment Re Component Cooling Water Sys for Reactor Coolant Pump Thermal Barriers, Per 900628 Meeting W/Nrc

ML18153C413
Person / Time
Site:	Surry
Issue date:	10/25/1990
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-442, NUDOCS 9011010027
Download: ML18153C413 (3)

Text

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e e VIRGINIA ELECTRIC AND PowER COMPANY RICHMOND, VIRGINIA 23261 October 25, 1990 United States Nuclear Regulatory Commission Serial No.90-442 Attention: Document Control Desk NO/ETS 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 REQUEST FOR ADDITIONAL INFORMATION RCP THERMAL BARRIER CCW DESIGN In a meeting with your staff of June 28, 1990, concerning Component Cooling Water system design for the RCP thermal barriers, two questions were posed by the staff that required additional information. These questions were also documented in your July 9, 1990, letter summarizing the meeting. The attachment to this letter provides the information necessary to complete your assessment.

• If you have any further questions or require additional information, please contact us.

Very truly yours, Jl~

W. L. Stewart Senior Vice President - Nuclear Attachment 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

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e ATTACHMENT QUESTIONS FROM JUNE 28, 1990 CCW THERMAL BARRIER MEETING WITH NRC

1. How will Virginia Power address post LOCA thermal expansion between the two new automatic containment isolation valves which has the potential to result in unacceptably high pressures between these two closed valves?

The new containment penetration will be provided with overpressure protection in accordance with ANSI/ANS-56.2-1984, "Containment Isolation Provisions for Fluid Systems After a LOCA." The containment isolation valve arrangement offers inherent overpressure protection. The inside isolation valve is designed so that trapped fluid applies pressure under the valve seat acting on the valve spring. The valve spring is set to allow the valve to open below the design pressure of the piping and valves (2485 psig). The Component Cooling Water (CCW) piping, upstream of the containment penetration from the check valve at the CCW inlet to the reactor coolant pump thermal barrier to the penetration, is protected from over pressurization by a thermal relief valve at each reactor coolant pump which will relieve at 2485 psig.

2. Will the new containment isolation valves and the associated new p1pmg be designed to the current regulatory requirements or to the original plant licensing basis?

The new containment penetration design consists of an existing six inch penetration which will function as a sleeve for the three inch component cooling water piping. The penetration will be modified by installing a closure plate on the inside of the containment. The new containment isolation valves, associated new piping and modified penetration will be designed to the original plant licensing basis. The piping passing through the penetration will be designed to USA Standard Code for Pressure Piping USAS 831.1.0-1967, "Power Piping." The containment isolation valves will be designed in accordance with the original plant

e design codes of ANSI 831.1.0-1967 and 816.5. If the valve supplier elects to furnish the valves to a later code, the later will be reconciled with the original plant design code as required by ASME XI, Article IWA-721 O(c). The modified penetration will be designed in accordance with the ASME Boiler and Pressure Vessel Code 1968, Section 111 and is designed to meet the stress limits provided in the UFSAR Section 15.5.1.8.