05000245/LER-1982-007, Forwards LER 82-007/03L-0.Detailed Event Analysis Encl

Forwards LER 82-007/03L-0.Detailed Event Analysis Encl

ML20050C848
Person / Time
Site:	Millstone
Issue date:	04/02/1982
From:	Mroczka E NORTHEAST UTILITIES
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20050C850	List: 05000245/LER-1982-007, Forwards LER 82-007/03L-0.Detailed Event Analysis Encl 05000245/LER-1982-007-03, /03L-0:on 820315,while Performing Isolation Condenser Isolation Instrument Functional & Calibr Test, Switch 1350B Tripped at 147 Inches of Water.Caused by Setpoint Drift.Switch Recalibr
References
MP-1-2040, NUDOCS 8204090406
Download: ML20050C848 (2)

LER-1982-007, Forwards LER 82-007/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2451982007R00 - NRC Website

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' 1 ' ' :7 April 2, 1982 MP-l-2040 Mr. Ronald C. Haynes Regional Administrator, Region 1 Office of Inspection and Enforcement D

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King of Prussia, Pennsylvania 19406

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Reference:

Provisional License DPR-21 u " g )gggy-Docket No. 50-245 (q3 c:p G {({3 ((a# }

1 Reportable Occurrence R0-82-7/3L

Dear Mr. Haynes:

7 This letter forwards the Licensee Event Report for Report $

" rrence R0-82-7/3L, required to be submitted within 30 days pursuant to the requirements of the Millstone Unit 1 Technical Specifications, Section 6.9.1.9.a.

An additional three copies of the report are enclosed.

Yours truly, NORTHEAST NUCLEAR ENERGY COMPANY

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E. J. Mroczka Station Superintendent tiillstone Nuclear Power Station EJM/TST:mo

Attachment:

LER R0-82-7/3L cc:

Director, Office of Inspection and Enforcement, Washington, D.C. (30)

Director, Office of Management Information and Program Control, Washington, D.C. (3)

U.S. Nuclear Regulatory Commission, c/o Document Management Branch, Washington, D.C.

20555 s

8204090406 820402 PDR ADOCK 05000245 PDR S

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ATTACHMENT TO LER 82-7/3L NORTHEAST NUCLEAR ENERGY COMPAY MILLSTONE NUCLEAR POWER STATION - UNIT 1 PROVISIONAL LICENSE NUMBER DPR-21 D0CKET NUMBER 50-245 Identification of Occurrence An Engineered Safeguard feature instrument setpoint was found to be less conservative than that established by Technical Specifications when the Isolation Condenser flow switch tripped outside of its allowable setpoint range.

Conditions Prior To Occurrence Prior to occurrence the unit was operating at a steady state power level of 100 percent.

Description of Occurrence On March 15,1982, at 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br />, while performing Isolation Condenser Isolation Instrument Functional and Calibration Test, switch 1350B tripped at 147 inches of water. Technical Specification, Table 3.2.1, requires automatic isolation of the Isolation Condenser when a line break on the steam supply reaches a differential pressure measuring greater than 150 inches of water but less than 164 inches of water.

Apparent Cause of Occurrence Failure of the switch in question to trip at its desired setpoint was attributable to setpoint drift.

Analysis of Occurrence The Isolation Condenser is a standby, high pressure system provided for the removal of fission-product decay heat after a reactor scram.

It is initiated on high reactor pressure in excess of 1085 psig sustained for 15 seconds or reactor 10-10 level.

It can also be manually initiated. The Isolation Con-denser operates by natural circulation, with steam flowing from the reactor through the condenser tubes, condensing and returning by gravity to the vessel via re-circulation loop 'B'.

There are two pressure switches on the steam supply line and the condensate return line that detect a line break by sensing differential pressure across

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pipe elbows located inside the drywell. When any one of the switches senses a high differential pressure (above Technical Specification setpoint limits) the Isolation Condenser is automatically isolated.

Failure of the switch to trip within its allowable setpoint range did not result in a condition that had not been previously analyzed.

If an actual line break had existed on the steam supply side automatic isolation of the Isolation Condenser would had occurred earlier than Technical Specification requirements.

In this case, emergency core cooling systems would be available to maintain reactor water level if reactor depressurization became necessary.

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Corrective Action

The pressure swit S in question was recalibrated to within its~ acceptable setpoint range and was satisfactorily tested.

The pressure switch is a Barton Model 288, with a range of 0-250 inches of water.