ML20070A924

Application for Amend to License NPF-86,changing RHR Suction Relief Valve Setpoint Upper Limit of Tech Specs 3.4.9.3 & 4.4.9.3.2 to Reflect Removal of RHR Isolation Valve Autoclosure Interlock

ML20070A924
Person / Time
Site:	Seabrook
Issue date:	01/24/1991
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20070A929	List: ML20070A924
References
NYN-91011, NUDOCS 9101290341
Download: ML20070A924 (3)

Text

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New Hampshire Ted C. Feigenbovm Yh hh President ond Chief Executive Ofhcer NYN 91011 January 24, 1991 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk

References:

(a) Facility Operating License No. NPF-86, Docket No. 50 443 (b) WCAP-11736 A, ' Residual llcat Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners Oroup,' Revision 0.0, October 1989

Subject:

Request for License Amendment Residual lleat Removal System isolation Valve Autoclosure Interlock Removal Gentlemen:

Pursuant to 10 CFR 50.90, New ilampshire Yankee (NIIY) hereby proposes to amend the Seabrook Station Operating License (Facility Operating License NPF 86) by incorporating the proposed changes, provided herein as Enclosure 1, into the Seabrook Station Technical Specifications. The proposed changes involve changes to the Residual lleat Removal (RilR) suction relief valve setpoint upper limit of Technical Specification 3.4.9.3, " Reactor Coolant System, Pressure / Temperature Limits, Overpressure Protection Systems', and the Surveillance Requirements of Technical Specifications 4.4.9.3.2,

• Reactor Coolant System, Pressure / Temperature Limits, Overpressure Protection Systems,' and 4.5.2, " Emergency Core Cooling Systems, ECCS Subsystems T., Greater Than or Equal to 350*F.' These changes are associated with the proposed removal of the Residual lleat Removal (RilR) Isolation Valve Autoclosure Interlock (ACl). The RilR ruction relief valve setpoint upper limit is decreased to ensure adequate relief valve capacity and overpressure protection. Revised 11ASES are provided as Enclosure 2 to this letter. In accordance with 10 CPR 50.36, these $

IIASES are not considered to be part of the Technical Specifications.

The Westinghouse Owners Group (WOG) undertook an effort to address concerns of the nuclear industry about inadver'ent RilR isolation events caused by spurious actuation of the ACI circuitry. The topical report resulting from this effort [WCAP-11736-A, Reference (b)] was reviewed and accepted by the NRC staff for use as a supplement for plant specific requests to remove the ACl. The Seabrook plant specific evaluation of the proposed changes is based on WCAP-11736 A and is provided as Enclosure 3.

Ilased upon the information contained in Enclosure 3, NHY has determined that the proposed changes do not constitute an unreviewed safety question pursuant to 10CFR50.59, nor do they involve a significant hazards consideration pursuant to 10CFR50.92, 9101290341 910124 I PDR ADOCK 05000443 jd0lO ,

P PDR D l

New Hampshire Yonkee Division of Public Service Company of New Hampshire -

i United States Nuclear Regulatory Commission January 24, 1991 Attention: Document Control Desk Page two New Hampshire Yankee has reviewed the proposed changes utilizing the criteria specified in 10CFR50.92 and has determined that the proposed changes would not

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of an accident previously evaluated in the PSAR is not increased since the WCAP 11736 A probabilistic analyses for Callaway are applicable to Scabrook. These analyses show that the net effect of ACI removal is an improvement in plant safety. ACI removal does not affect an) function of the RilR system or Reactor Coolant System (RCS) other than automatic isolation. The ACI function was provided to ensure that the Isolation valves are closed during heatup to prevent an intersystems LOCA.

Adequate RHR system protection is provided by administrative controls, alarms, and the RHR suction relief valves. Reducing the RHR suction relief valve setpoint upper limit ensures adequate relief capacity and overpressure protection.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated. ACI removal does not change the function or failure modes of the RIIR suction isolation valves that could cause an accident of a different kind. The change in the RilR suction relief valve setpoint upper limit ,'ill not affect the passive function of the valves and ensures that the required overpressure protection is provided.

3. Involve a significant reduction in a margin of safety. Reducing the RHR suction relief valve setpoint upper limit ensures adequate relief capacity and overpressure protection in the event of an overpressure transient. ACI removal eliminates one of the possible causes of spurious valve closure that could reduce the availability of the RHR suction relief valves for RCS low temperature overpressure protection. Also, the probabilistic analyses demonstrate that administrative controls, combined with the valve not closed alarms, increases the probability of all RHR suction valves being closed when the RCS is at high pressure thereby providing greater assurance that the ECCS subsystems are properly aligned and reducing the probability of an intersystems LOCA.

Enclosure 4 provides the NilY responses to the NRC staff positions contained in Section 3.0 of the staff's safety evaluation of WCAP 11736.

Implementation of the modifications required to delete the RHR autoclosure interlock are scheduled to occur during the next refueling outage. New Hampshire Yankee requests approval of these proposed changes by May 31,1991 to support this implementation schedule.

Should you have any questions regarding this request, please contact Mr. Terry L.

Harpster, Director of Licensing Services, at (603) 474 9521, extension 2765.

Very truly yours, h &$ #

Ted C. Feigenbaum

United States N lear Regulatory Commission January 24, 1991 Attention: Doc nt Control Desk Page three 4

cc: Mr. Thomas T. Martin Regional Administrator United States Nuc! car Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Mr. George L. lverson, Director Office of Emergency Management State Office Park South ,

107 Pleasant Street Concord, Nil 03301 Mr. Gordon Edison, Sr. Project Mgr.

Project Directorate 13 i Division of Reactor Projects l

U.S. Nuclear Regulatory Contmission '

Washington, DC 20555 Mr. Noel Dudicy l NRC Senior Resident inspector P.O. Box 1149 -

Scabrook, NH 03874 Enclosure (s)

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