05000423/LER-1997-031, :on 970507,RHR Valve Low Pressure Open Permissive Bistable Setting Was Set non-conservatively. Caused by Implementation of Incorrect Calibration Info. Re-evaluated RHR Suction/Isolation Valve LPI Calibrations

:on 970507,RHR Valve Low Pressure Open Permissive Bistable Setting Was Set non-conservatively. Caused by Implementation of Incorrect Calibration Info. Re-evaluated RHR Suction/Isolation Valve LPI Calibrations

ML20140G601
Person / Time
Site:	Millstone
Issue date:	06/05/1997
From:	Danni Smith NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20140G594	List: 05000423/LER-1997-031, :on 970507,RHR Valve Low Pressure Open Permissive Bistable Setting Was Set non-conservatively. Caused by Implementation of Incorrect Calibration Info. Re-evaluated RHR Suction/Isolation Valve LPI Calibrations
References
LER-97-031, LER-97-31, NUDOCS 9706160365
Download: ML20140G601 (3)

LER-1997-031, on 970507,RHR Valve Low Pressure Open Permissive Bistable Setting Was Set non-conservatively. Caused by Implementation of Incorrect Calibration Info. Re-evaluated RHR Suction/Isolation Valve LPI Calibrations

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
4231997031R00 - NRC Website

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NRC FORM 366 U.s. NUCLEAR REGULATORY CoMMISsloN APPROVED BY CpMS NO.3160-0104 (4-96)

EXPIRES 04/30/98

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FACILITY NAME (1)

DOCKET NUMSER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 3 TITLE M)

RHR Valve Low Pressure Open Permissive Bistable Setting Set Non-Conservatively EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVisloN MONTH DAY YEAR FACIUTY NAME DOCKET NUMBER NUMBER NUMBER f

05 07 97 97 031 00 06 05 97 OPERATING 5

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTSoF 10 CFR 1: (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v)

X so 73(a)(2)(ii So.73(a)(2)(viii)

POWER 000 20.2203(a)(1) 20.2203(a)(3)(i) 50.73!$(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(si) 50.73(a)(2)(iii) 73.71

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20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv) oTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(V)

Specify in Abstract tylow l

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20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FoR THIS LER (12)

NAME TELEPHONE NUMBER tinclude Area Codel David A. Smith, MP3 Nuclear Licensing Manager (860)437-5840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT M A NUF ACTURE R REPORTABLE J

TO NPRDS TO NPROS q

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR No SUBMISSloN

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YES DATE (15)

(if yes, complete EXPECTED sUBMtssioN DATE).

ABSTRACT (Limit to 1400 spaces,i.e., approximately15 single-spacedtypewnttenlines) (16)

At 1330 on May 7,1997, with the Unit in Mode 5, a system engineering review of the Reactor Coolant System (RCS) l wide range pressure channel calibration procedures concluded that the Residual Heat Removal System (RHR) Low Pressure Interlock (LPI) setpoint did not comply with Technical Specification (TS) Surveillance Requirement 4.5.2.d.1.

The existing LPl setting allows the RHR isolation valves to be opened at RCS pressures higher than those presented in ths TS. Consequently, this event is reportable pursuant to 10CFR50.73(a)(2)(i)(B), as a condition or operation prohibited by the plant's TS.

l-l This condition was caused by the implementation of incorrect calibration information supplied during unit initial startup.

However, there were no adverse consequences as a result of the event since the RHR pressures did not exceed the RHR suction header relief valve limits.

4 No immediate corrective action is required, however, prior to entry into Mode 4, the RHR suction / isolation valve LPI bistable calibrations will be re-evaluated and the LPl bistables will be recalibrated to comply with the TS.

9706160365 970605 PDR ADOCK 05000423 S

PDR NRC FORM 366 (4 05) l

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lU.s. NUCLEAR REGULATORY Commission l

(4 95) i UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION l

FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVislON Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 3 97 031 00 TEXT (if rnore space is required, use additionalcopies of NRC Forrn 366A) t17) l l

l l.

Description of Event

At 1330 on May 7,1997, with the Unit in Mode 5, a system engineering review of the Reactor Coolant System (RCS)

I wids range pressure channel calibration procedures concluded that the Residual Heat Removal System (RHR) Low l

Pressure Interlock (LPI) setpoint did not comply with Technical Specification (TS) Surveillance Requirement 4.5.2.d.1.

TS 4.5.2 states that *Each ECCS subsystem shall be demonstrated OPERABLE: d) At least once each REFUELING INTERVAL by: 1) Verifying automatic interlock action of the RHR System from the actual Reactor Coolant System (RCS) by ensuring that with a simulated or actual RCS pressure signal greater than or equal to 390 psia the interlocks prevent the valves from being opened." The unit procedure functionally tests the LPls once every refueling interval by simulating a 500 psia high pressure signal and calibrates the bistables between 390 and 403 psia.

Th3 existing LPI bistables were first calibrated in March 1985 dunng preoperational testing of the RCS instrumentation using calibration information supplied in Loop Calibration Reports and in accordance with approved plant procedures.

B:cause of incorrect information, the bistable was configured to trip on decreasing pressure instead of increasing pressure as required by the TS.

As a result of the incorrect calibration, the LPIs prematurely trip on decreasing pressure (403 psia) and reset on increasing pressure such that the RHR isolation valves can be opened with RCS pressure as high as 433 psia. This is above the 390 psia TS limit and is reportable pursuant to 10CFR50.73(a)(2)(i)(B), as a condition or operation prohibited by the plant's technical specifications.

II.

Cause of Event

This condition was caused by the implementation of incorrect calibration information supplied during unit startup.

Ill. Analysis of Event Th3 present LPI setpoint would potentially allow opening of the RHR suction / isolation valves at pressures as high as 433 psia. However, a review of the equipment history for the RHR suction line relief valves (3RHR*RV8708A&B) has shown that the RHR header relief valves have not lifted (requiring reset) which would indicate that pressures on the suction side of the RHR system have remained below the 440 psig [455 psia](ref. LER 96-034-00) lifting setpoint and 600 psig [615 psia) design limits. In addition, RHR system overpressure prevention is administratively controlled in the cp; rations RCS cool-down procedure which requires that RCS pressure be less than 390 psia before placing RHR in operation. Consequently, even though the violation of the TS surveillance requirement occurred, there were no cdverse consequences as a result of this event.

IV. Corrective Action

j No immediate corrective action is required, however, prior to entry into Mode 4:

1.

The Residual Heat Removal System suction / isolation valve Low Pressure Interlock bistable calibration will be re-evaluated and the low pressure permissive bistable will be recalibrated to comply with the TS.

l l

l

.U.S. NUCLEAR REGULATORY Commission 84 95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 3 97 031 00 TEXT (If more space is required, use additionalcopies of NRC Form 366A) (17)

V.

Additional Information

None

Similar Events

LER 96-034-00 Residual Heat Removal System (RHR) Pump Suction Relief Valve Setooint Not in Accordance With

- Technical Specifications (TS)

The TS require that the RHR pump suction relief valves be set.450 psig to provide adequate over pressure protection when the temperature of any Reactor Cools nt System cold leg is less then 350 degrees Fahrenheit. By recommendation of the A/E and contrary to the TS requirement, the actuallift pressure for the RHR pump suction relief was revised to 440 psig without issuing a TS change.

Manufacturer Data Ells System Code Residual Heat Removal..

..BP Ells Component Code Vel v e, I s o l a t i o n.................................................................. l S LV