ML20245A605

Forwards Pressurizer Code Safety Valve Reliability, Technical Review Rept,Initiated Due to LER 86-018.AEOD Should Initiate Case Study to Determine Extent of Problems & to Assess Adequacy of Efforts Toward Increasing Reliability

ML20245A605
Person / Time
Site:	Diablo Canyon
Issue date:	03/27/1987
From:	Wegner M NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	Lam P NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
Shared Package
ML20245A607	List: ML20245A605 ML20245B510
References
NUDOCS 8704270034
Download: ML20245A605 (1)

Text

. _ - . __

? % a "o,, UNITED STATES ,

g NUCLEAR REGULATORY COMMISSION {

$ r WASHINGTON, D. C. 20666

%,*****'/ g g7 llB7 i

i MEMORANDUM FOR: Peter Lam, Chief 1 Reactor Systems Section 2 Reactor Operation Analysis Branch Office for Analysis and Evaluation of Operational Data FROM: Mary S. Wegner, Reactor Systems Engineer Reactor Systems Section 2 Reactor Operations Analysis Branch Office for Analysis and Evaluation of Operational Data

SUBJECT:

TECHNICAL REVIEW REPORT: PRESSURIZER CODE SAFETY VALVE RELIABILITY The subject technical review report is enclosed for your consideration. The study was initiated as a result of the screening of a Diablo Canyon 1 licensee event report (LER 86-018). Our evaluation found that upon testing their pressurizer Code safety valves with the unit at hot standby, Diablo Canyon 1 personnel found their lift setpoints to be above their technical specifications (TS) limits. The test method was questioned and its proper use was determined.

The valves were correctly reset. A data search indicated that a total of 34 pressurizer Code safety valves at 17 plants had setpoint drift, leakage, misadjusted ring settings, or maintenance / installation problems since January 1, 1983. These problems could lead to inadvertent reactor scrams, overpressure-ization of the reactor coolant system (RCS), or degradation of the reactor coolant pressure boundary. These problems appear to be generic to all safety /

relief valves. It is suggested that AEOD initiate a case study to determine the extent of the problems and to assess the adequacy of present efforts toward increasing safety valve reliability.

. vary S. We er, Reactor Systems Engineer Reactor Sy tems Section 2 Reactor Operations Analysis Branch Office for Analysis and Evaluation of Operational Data

Enclosure:

As stated cc w/ enclosure:

T. Gwynn, OCM i. Gifford, GE s L. Brinkman, CE W. Beaumont, W l O-f R. Borsum, B&W t en--o