05000219/LER-1983-005, Forwards LER 83-005/03L-0.Detailed Event Analysis Encl

Forwards LER 83-005/03L-0.Detailed Event Analysis Encl

ML20069A692
Person / Time
Site:	Oyster Creek
Issue date:	03/04/1983
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20069A694	List: 05000219/LER-1983-005, Forwards LER 83-005/03L-0.Detailed Event Analysis Encl
References
NUDOCS 8303160197
Download: ML20069A692 (3)

LER-2083-005, Forwards LER 83-005/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2192083005R00 - NRC Website

text

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GPU Nuclear

~ NQQIQf P.O. Box 388 Forked River, New Jersey 08731 609-693-6000 Writer's Direct Dial Number:

March 4, 1983 Mr. Ronald C. Haynes, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406

Dear Mr. Haynes:

Subj ect: Oyster Creek Nuclear Generating Station Docket No. 50-219 Licensee Event Report Reportable Occurrence No. 50-219/83-05/03L This letter forwards three copies of a Licensee Event Report (LER) to report Reportable Occurrence No. 50-219/83-05/03L in compliance with paragraph 6.9.2.b.2 of the Technical Specifications. We realize this LER is being submitted beyond the time limitation specified in Technical Specifications, paragraph 6.9.2.b. The cause of the delay is attributed to administrative delay within the department responsible for the investigation of the event described herein and the preparation of this LER.

Very truly yours, Peter B. Fiedler Vice President and Director Oyster Creek PBF:jal Enclosure s cc: Director (40 copies)

Of fice of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dire ctor (3)

Of fice of Management Information and Program Control U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC Resident Inspector Oyster Creek Nuclear Generating Station y Forked River, NJ 08731 f-8303160197 930304 clear is a part of the General Pubhc Utihties System /

PDR ADOCK 05000219 g PDR

1 OYSTER CREEK NUCLEAR GENERATING STATION Forked River, New Jersey 08731 Licensee Event Report -

Reportable Occurrence No. 50-219/83-05/03L Report Date March 4, 1983 Occ ur re nc e Date January 26, 1983 Identification of Occurrence Containment Spray System high drywell pressure switches IP15A, IP15B, and IP15C tripped at a value greater than that specified in the Technical Specifications, Table 3.1.1, Item E.1.

This event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9.2.b(2).

Conditions Prior to Occurrence The plant was operating at steady state power. Plant parameters at the time of occurrence were:

Mode Switch Run Thermal Power 917 MWt Generator Load 244 MWe Reactor Coolant Temp. 5370F Description of Occurrence On Wednesday, January 26, 1983, while performing the " Containment Spray System Automatic Actuation Test," procedure 607.3.002, the IP15A, IP15B, and IP15C trip points were found to be less conservative than those specified in the Technical Specifications. Surveillance testing on the High Drywell Pressure Switches for the Containment Spray System revealed the following data:

Pressure Switch Designation Desired Setpoint As Found As Left IP15A 2.0 psig 2.15 psig 1.84 psig IP15B 2.0 psig 2.24 psig 1.84 psig IP15C 2.0 psig 2.10 psig 1.80 psig IP15D 2.0 psig 1.91 psig 1.91 psig

- ._ -. =- . - _ - ._ . - - - - - _ - - _ ..

Licensee Event Report Page 2 Reportable Occurrence No. 50-219/83-05/03L Apparent Cause of Occurrence The cause of the occurrence has not yet been determined. After reviewing previous survie11ance data sheets from the past 12 surviellances, it has been determined that the setpoint drif t found in this surveillance is not indicative of previous setpoint drif t. Further investigation will be required to determine the cause of this unusally high instrument drif t.

Analysis of Occurrence The Con *.ainment Spray System consists of two independent cooling loops, each capable of removing fission product decay heat from the primary containment af ter a postulated loss of cooling accident. The containment spray system automatically actuates upon receipt of two high drywell pressure and two reactor low-low water level signals in either of two trip systems.

This function would have been delayed by approximately 0.1 seconds from the start of a Design Basis Accident. Additionally, since the reactor low-low water level setpoint is not expected until approximately 4 seconds from the start of the design basis accident , the delay in actuating the high drywell pressure switches would have no ef fect on initiating the containment spray system.

The safety significance of this event is considered minimal since the high drywell pressure switches would have actuated but at a slightly higher pressure than the required setpoint.

i Corrective Action Pressure switches IP15A, IP15B, and IP15C were adjusted to trip within the Technical Specification limit of 2.0 psig. Investigation is continuing to determine the cau,se of the instrument setpoint drif t. Additionally, procedure acceptance criteria will be reviewed to possibly accommodate problems encountered from instrument repeatability.

Failure Data Ma nufacturer : ITT Barton Model: 228A Indicating Pressure Switch Range: 0-10 psig