ML20045F774

LER 93-001-01:on 930305,cooldown Exceeding Limits of TS 3.4.9.1 Experienced After Switching from SG Cooling to Dh Sys Cooling.Caused by Failure of Cv Controller.Valve Repaired & Valve Operation Instructions Revised

ML20045F774
Person / Time
Site:	Crystal River
Issue date:	06/30/1993
From:	Stephenson W FLORIDA POWER CORP.
To:
Shared Package
ML20045F773	List: ML20045F772 ML20045F774
References
LER-93-001, LER-93-1, NUDOCS 9307080392
Download: ML20045F774 (4)

Text

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OF M ANAGEMENT AND BUDGET, WASHINGTON DC 20603 FACILITY NAME (1)

DOCKET NUMBER (2) PAGE (3)

CRYSTAL RIVER UNIT 3 (CR-3) oj 5l 0l ol ol 3l ol 2 TITLE (4) 1 lOFl 0 l 4 Failure of Decay Heat System Cooling Control Valve Causes Cooldown Exceeding Technical Specification Limits EVENT DATE(5) LER NUMBER (6) REPOHT DATE (7)

OTHER F AC8LITIES NNOLVED (8)

SEQUENTIAL REVISION F ACILITY NAMES DOCKET NUMBER (S)

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TELEPHONE NUMBER ARE.A CODE W. A. Stephenson, Nuclear Safety Supervisor 9l0l4 7l 9 l 5 l- l 6l 4 l 8 l 6 AUSE SYSTEM COMPLETE ONE LINE FOR EACH COMPONENT FAILURE IN TthS HEPOfIT (13)

COMPONENT MANUF AC. FEPORTABLE CAUSE SYSTEU COMPONENT MANUFAC- '

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EXPECTED MONTH DAY YEAR G YES SUBMISS 80N JNm conumer LKPECTED SUBwSSCN WE) NO w si m CT ,L - ., DATE 06)

-+,, w -- .- e e3 l l l On March 5,1993, Crystal River Unit 3 (CR-3) was in Mode 4 (Hot Shutdown) and cooling down for a planned maintenance outage. Two reactor coolant system (RCS) pumps were operating and the RCS was being cooled by the steam generators.

After switching from steam generator cooling to Decay Heat (DH) system cooling, a cooldown exceeding the limits of Technical Specification 3.4.9.1 was experienced.

Af ter securing the RCS pumps, operators, unable to establish control of the DH cooling system from the control room, dispatched an auxiliary building operator to establish manual control. While using posted instructions to convert from automatic to manual control, a valve inadvertently opened causing additional RCS cooldown. The valve was immediately closed manually. After several minutes, the valve started to drift open due to damage to a stem key connecting the valve to the manual handwheel. The alternate DH train was then placed in operation, stabilizing RCS temperature.

The initial excessive cooling was caused by failure of a control valve controller. Additional cooling was caused by inadequate posted instructions.

Damage to the valve stem key was caused by improper manual operation. The valve was repaired and revised valve operation instructions were posted. An evaluation have been of this event has been conducted and additional corrective actions identified.

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VENT DESCRIPTION:

On March 5,1993, Crystal River Unit 3 (CR-3) was in Mode 4 (Hot Shutdown) and cooling down for a planned maintenance outage. Two reactor coolant system (RCS) pumps [AB,P] were operating and the RCS was being cooled by the steam generators

[AB,HX). The RCS temperature was 263 degrees Fahrenheit (*F). RCS pressure had been reduced to 200 pounds per square inch gauge (psig) to allow the operators to place the Decay Heat (DH)[BP] system in service. While switching from steam generator cooling to decay heat system cooling, a cooldown that er.ceeded the limits l, of Technical Specification (TS) 3.4.9.1 was experienced.

At approximately 1239, the "A" DH train was placed in service with the two RCS pumps still in operation. The RC pumps were secured at 1245 after assuring DH system operation by verifying a drop in DH system outlet temperature. At this time, RCS cold leg temperature had dropped to 256*F.

l When switching the mode of core cooling, the temperature monitoring point for the

! reactor vessel wall changes. With RCS pumps on, the bulk temperature of the RCS, as measured by cold leg temperature instruments [AB,TI], is an accurate measure of the actual temperature of the vessel wall. When RCS pumps are secured and DH is l

providing core cooling, the DH heat exchanger [BP,HX] outlet temperature is used.

! This is necessary because DH cooled RCS fluid is injected directly into the reactor vessel without the benefit of mixing with the bulk RCS fluid. At the time that the RCS Pumps were secured, the DH heat exchanger outlet temperature was 229'F.

After securing the RCS pumps, control room licensed operators, using the DH system temperature controller [CC,TC], attempted to control the cooling from the DH system. After several attempts to reduce the rate of cooling, an auxiliary building non-licensed operator was dispatched to the cooling water control valve

[CC, TCV] providing cooling water to the DH system heat exchanger, DCV-177, to l manually close it. At this time, the DH cooler outlet temperature was 210*F. l l

At 1322, while converting the cooling control valve, DCV-177, from automatic control to manual control, the valve inadvertently opened causing an additional RCS cooldown. The auxiliary building operator immediately closed the valve. Over the next several minutes, the valve started to drift open. At 1330, the DH cooler outlet temperature was 143*F.

At this time, it was apparent to the control room operators that the cooldown had l not been stopped by manually closing the cooling water control valve. At 1331, the alternate "B" DH train was started and several minutes later the "A" DH train was secured. Starting the alternate DH train caused an additional brief DH cooler outlet temperature step change to 70*F, the ambient temperature of the alternate train.

NED Form 366A (6-69)

NFC FCM 306A U.S. NUCLEAR IGOULATC]tY COMMISSKJN APPF:OVED OMB NO.3160 4 100 EXPtRES C/IDC2 LICENSEE EVENT REPORT (LER) (sT g T,E,Dg u g a gsPONg TgC gou gTg  ;

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. sEQUEN76AL HEV$CN YEAR WWBER IWMBER CRYSTAL RIVER UNIT 3 (CR-3) 0l 5] 0l 0l 0l 3l 0l 2 9l3 -

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0l1 0 l 3 lOFl 0 l 4 text g . - u -ou . (m Using the B" DH train, the DH cooler outlet temperature was stabilized at 220*F.

RCS temperature was maintained at this temperature for the next three hours to allow any stresses induced by the cooldown to diminish.

CAUSE:

Technical Specifications permit a step change of 25*F when transitioning to DH.

The initial overcooling was approximately 25*F greater than permitted by TS and was '

caused by failure of the controller for control valve DCV-177 that directs cooling water through the DH system heat exchanger. The cause of the automatic control failure has been determined to be a lack of timely preventive maintenance resulting in debris in the electro-magnetic transducer controlling DCV-177.

The manual failure was caused by improper operation of the valve after an unsuccessful attempt to follow the locally posted instructions. The operator was unable to remove, as required by the instructions, a linkage pin connecting the valve positioner to the manual handwheel. This hindrance was probably due to corrosion on the pin. The operator then chose to manually close the valve from the full open position with the actuator attached. While this initially closed the valve, it ultimately overstressed the stem key which freed the stem from the handwheel and allowed the valve to reopen. Had the linkage pin been capable of removal as required, the action to close the valve would have terminated the overcooling event.

The instructions for assuming manual control contributed to the event. The initial instruction caused the valve to reposition to full open resulting in additional overcooling.

EVENT EVALUATION:

Reactor vessel cooldown limits are provided to assure analysis assumptions used to calculate the RCS pressure / temperature limits are not exceeded. The pressure / temperature limits, included in the Technical Specifications, assure that stresses induced by system pressure and thermal gradients across the vessel wall do not exceed the stress limits for cyclic operation. The calculation of these limits are based on RCS fracture toughness properties. CR-3 has completed an engineering evaluation of this overcooling and has determined the effects of this cooldown on the fracture toughness properties of the RCS were negligible.

CORRECTIVE ACTIONS:

The control valve controller has been repaired. Additionally, the valve, DCV-177, has been repaired. A failure analysis has been completed and has identified further corrective actions. Placards cositaining revised interim instructions have been installed for DCV-177 and similar valves servicing the DH coolers. The most efficient sequence of steps for taking manual control of the valve is still being NICC F orm 366A (6-89)

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CRYSTAL river UNIT 3 (CR-3) 0l 6l 0l 0l 0l 3l 0l 2 9l3 -

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0l1 0 l 4 lOFl 0 l 4 TEXT rst more erece e nee.ned use amar ,w mac Form asu e (17) determined. The procedure addressing manual valve operations will be appropriately revised. Preventive maintenance practices will be improved for these components.

Additional corrective actions included evaluating procedure changes for placing a DH train in service and addressing the temperature consequences of placing an alternate DH trait, in operation.

PREVIOUS SIMILAR EVENTS:

There have been no previous events involving a reactor vessel cooldown exceeding the TS limits.

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NRC Form 366A (6-89)