05000219/LER-1981-064, Forwards LER 81-064/03L-0.Detailed Event Analysis Encl

Forwards LER 81-064/03L-0.Detailed Event Analysis Encl

ML20041A762
Person / Time
Site:	Oyster Creek
Issue date:	12/14/1981
From:	Carroll J JERSEY CENTRAL POWER & LIGHT CO.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20041A758	List: 05000219/LER-1981-064, Forwards LER 81-064/03L-0.Detailed Event Analysis Encl 05000219/LER-1981-064-03, /03L-0:on 811203,containment Spray High Drywell Pressure Switches IP-15A,IP-15B,IP-15C & IP-15D Found to Trip at Values Greater than Limit.Caused by Instrument Repeatability.Switches Reset
References
NUDOCS 8202220476
Download: ML20041A762 (3)

LER-1981-064, Forwards LER 81-064/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2191981064R00 - NRC Website

text

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CYSTER CREEK NUCLEAR GENERATING STATION b

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(609) 693-1951 P.O. BOX 388

• FORKED RIVER

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December 14, 1981 Mr. Ibnald Haynes, Director Office of Inspection and Enforcemnt Region I United States Nuclear Regulatory Cm mission 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. Haynes:

SUBJECT: Oyster Creek Nuclear Generating Station Docket No. 50-219 Licensee Event Report Reportable Occurrence No. 50-219/81-64/3L This letter forwards three copies of a Licensee Event Report to report Reportable Occurrence No. 50-219/81-64/3L in compliance with paragraph 6.9.2.b.1 of the Technical Specifications.

Very truly yours,

/

0.4

. T. Carroll, J Acting Director Oyster Creek Enclosures cc: Director (40 copies)

Office of Inspection and Enforcemnt United States Nuclear Regulatory Cmmission Washington, D.C.

20555 Director (3)

Office of Managenent Information and Program Control Unital States Nuclear Regulatory Cmmission Washington, D. C. 20555 NRC Resident Inspector (1)

Oyster Creek Nuclear Generating Station Forked River, N. J.

8202220476 820210 p#

PDR ADOCK 05000 e

r OYSTER CREEK NUCLEAR GENERATING STATICN Forked River, New Jersey 08731 Licensee Event Report Reportable Occurrence No. 50-219/81-64/3L Report Date Deceber 14, 1981 Occurrence Date Deceber 3, 1981 Identification of occurrence The Containment Spray high drywell pressure switches IP-15A, IP-15B, IP-15C, and IP-15D tripped at a value greater than the limit given in the Technical Specifica-tions Table 3.1.1 its E.1.

This event is considered to be a reportable occurrenm as defined in the Technical Specifications, paragraph 6.9.2.b.l.

Conditions Prior to Occurrence The plant was operating at steady state power.

Major Plant Parameters power:

Reactor 1859 Nht Generator 650 IWe 4

Flow:

Recirculation 15.4 x 10 gpn Feedwater 6.94 x 106 lb/hr Description of Occurrence On Thursday, Dec eber 3, at 4:00 P.M., while performing the " Containment Spray i

System Autmatic Actuation Test", the IP-15A, IP-15B, IP-15C and IP-15D trip l

points were found to exceed the Technical Specification limits. Surveillance testing yielded the following data:

Desired Switch Setpoint (psig)

As Found (psig)

As Left (psig)

IP-15A

<2 2.1 1.95 IP-15B 72 2.02 1.99 IP-15C 72 2.03 2.00 IP-15D 72 2.03 1.94 L.

Reportable Occurrence Page 2 Report No. 50-219/81-64/3L Apparent Cause of Occurrence The cause of the occurrence was attributed to instrument repeatability. 'Ihe switches were last reset to trip at 1.98, 1.92, 1.94 and 1.95 respectively. The greatest setpoint change of the four is 0.12 psig, which is well within the 0.20 psig - 0.30 psig design range of repeatability.

Analysis of Occurrence The Containment Spray Systs is made up of two independent cooling loops, each of which is capable of reoving heat frm the primary containment in the evet of a loss of coolant accident. In order to initiate the Containment Spray Systs, a cmbination of a high drywell pressure signal plus a reactor low-law water level signal must be received.

Although the high drywell pressure switches would have tripped at a slightly higher pressure than the desired setpoint, their actuation would have only been delayed a fraction of a second (approximately 0.1 second frcm the start of the design basis loss of coolant accident). However, the reactor low-low water level setpoint is not reached until almost 4 seconds frm the start'of the design basis loss of coolant accident. Due to this fact, the delay in actuation of the pressure switches auld not have affected the initiation of the Containment Spray Syst s, and, therefore, the safety significance of the occurrence is considered minimal.

Corrective Action

Pressure switches IP-15B and IP-15D were reset to trip within the Technical Specifications limit of 2.0 psig (as shown in the "As Icft" values in the Description of Occurrence). The drift proble of these snap-action switches is being investigated, along with possible setpoint changes to account for instrument repeatability. In addition, it has been reccmnended that these type of snap-action switches be replaced with an improved model and it is planned to replace thm during a future refueling outage.

Failure Data Manufacturer: ITT Barton Model:

228A Indicating Pressure Switch Range:

0-10 psig