05000312/LER-1978-006, Forwards LER 78-006/03L-0

Forwards LER 78-006/03L-0

ML20147B245
Person / Time
Site:	Rancho Seco
Issue date:	07/05/1978
From:	Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
Shared Package
ML20147B249	List: 05000312/LER-1978-006, Forwards LER 78-006/03L-0 05000312/LER-1978-006-03, /03L-0:on 780611,during Normal Operation,Increased Leakage Noticed,Between 1.1-1.7 Gallons/Min.Caused by Leaking Pressure Safety Valves & Block Valve on Electromatic Relief Valve.Valves Replaced or Reworked
References
NUDOCS 7810100050
Download: ML20147B245 (2)

LER-1978-006, Forwards LER 78-006/03L-0

Event date:
Report date:
3121978006R00 - NRC Website

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esuun SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacramento, California 95813; (916) 45b3211 July 5, 1978

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in Nuclear Regulatory Commission pg C

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ATTN:

Mr. R.11. Engelken

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NRC Operations Offlee, Region V gj w

Q 1990 N. Califernia Boulevard

. Walnut Creek Plaza, Suite 202 xrw U1' Da kSi5 Diij Walnut Creek, California 94596

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Re: Operating License DPR-54 j

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Reportable Occurrence 78-6

Dear Mr. Engelken:

In accordance with Technical Specifications for Rancho Seco Nuclear Generating Station, Section 6.9.4.2 h, the Sacramento Municipal Utility District is hereby submitting a thirty day report of Reportable Occurrence 78-6.

During the latter part of May, 1978, leak rate calculations performed per SP 200.02, the Daily Instrument Surveillance and SP 207.04, Evaluating RCS Leakage, although within the 1 gpm unidentified leak rate per Technical Specifications Section 3.1.6, showed an increase. This was attributed to known Icakaga from the pressurizer c1cetromatic relief, PV-21511, its assoc-lated block valve HV-21505, and one of the pressurizer code safety valves, PV-21506.

On Sunday, May 21, 1978, a thorough Reactor Building inspection was made looking for leaks. This reinforced previous findings that the major source of leakage was from the above mentioned valves.

PV-21506 was deter-mined to have a ruptured bellows, HV-21505 was determined to be Icaking past its seat, and PV-21511 appeared to be leaking past its pilot valve. A rep-resentative from the Dresser Valve and Instrument Division of Dresser Indust-ries was present to observe the leaking valves.

Based on the inspection and discussion'with the Dresser Industries representative, it was determined that there was no safety problem relating to the leaking valves.

Since the leakage was identifie.d, the majority of the leakage was to the pressurizer relief tank, and the valve leakage was determined not to pre-1 sent a safety problem, Technical Specifications Section 3.1.6.2 was not con-sidered applicable.

An engineering evaluation of the leakage from the valves, which was not j

going to the pressurizer relief tcnk but rather to the building atmosphere, i

and then showing up in the Reactor Building Drain accumulation tank was done.

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$00k;6 g llO fl AN ILlCTRtC SYSTEf4 S E RVIN G MDRE THAN 600,000 IN THE HLART C A l l f 0 F: h I A

b' R.11. Engelken July 5, 1978 Utilizing this evaluation and review by the Plant Review Committee, Meeting No. 502, a conservatively low 0.56 gpm value was assigned to this now considered identified leakage.

This value was used in subsequent leak rate calculations per SP 207.04.

Basically this allowed reindexing the leak rate procedure to show any new leak from the RCS.

It was also decided that if the leak rate calculation exceeded the 1 gpm unidentified leak rate w511e utilizing the 0.56 gpm value as identified, another inspection and engineering evaluation would be necessary.

On June 11, 1978, the calculated leak rate per SP 207.04 exceeded the 1 gpm unidentified.

Follow-up calculations utilizing the same procedure indicated leak rates varying from 1.1 gpm to 1.7 gpm.

On June 12, 1978, within twenty-four hours of the first unacceptable leak rate, the reactor was shut down per Technical Specifications Section 3.1.6.2.

Rather than conduct another engineering evaluation and determine a new value for the identified leakage f rom the volves, it was decided that repair and replace-ment of the problem valves was a more logical course of action. The known leaking code safety valve PV-21506 and the electromatic relief valve PV-21511 were replaced. The electromatic relief block valve HV-21505 was reworked (disc / seat lapped) and the other code safety valve PV-21507 was completely rebuilt and all new internal parts installed. The replacement valves were identical to the originals.

The three relief valves were bench-tested prior to installation and hydro-assist tested with the system pressure at 1500 PSI.

The three relief valves were all manufactured by Dresser Valve and Instrument Division of Dresser Industries (PV-21506 and PV-21507 being Consolidated Closed Bonnet Maxiflow Safety Valves, 3 inch, and PV-21511 being a Consolidated Elec tromatic Relief Valve, 2-1/2 inches). The block valve HV-21505 is a 2-1/2 inch gate valve manufactured by Velan Valve Corporation.

Respectfully submitted, A

Y. J. Sbttimoe J

Assistant General Manager and Chief Engineer JJM:Hil:jlk Attachment es: Director, MIPC (3)

Director, IE (30)