ML20043F507

LER 90-009-00:on 900512,feedwater Isolation Occurred as Result of Steam Generator 1B Reaching hi-hi Level Setpoint of 82% Level.Caused by Inappropriate Action Because of Lack of Attention to Detail.Feedwater Logic reset.W/900611 Ltr

ML20043F507
Person / Time
Site:	McGuire
Issue date:	06/11/1990
From:	Mcconnell T, Sipe A DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-009-01, LER-90-9-1, NUDOCS 9006150073
Download: ML20043F507 (7)

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!!!00 Hagen Ferry kbad Huntentille, NC28018 8985 DUKE POWER June 11, 1990 U.S. Nuclear Regulatory Commission Document Control Desk t Washington, D.C. 20555

Subject:

McGuire Nuclear Station Unit 1 Docket No. 50-369  !

Licensee Event Report 369/90-09 Gentlemen:

Pursuant to'10 CFR 50.73 Sections (a)(1) and (d), attached is Licensee Event Report 369/90-09 concerning a feedwater isolation as a result of Steam Generator 1B reaching its Hi-Hi-level setpoint while in Mode 5. This report is being submitted in accordance with 10 CFR 50.73(a)(2)(iv). This event is considered to be of no significance with respect to the health and safety of the public.

Very ruly you,s,

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, p T.L. McConnell DVE/ADJ/cb1 l Attachment xct- 'Mr. S.D. Ebneter American Nuclear Insurers l

Administrator, Region II c/o Dottie Sherman, ANI Library U.S. Nuclear Regulatory Commission The Exchange, Suit 245 101 Marietta St., NW, Suite 2900 270 Farmington Avenue l

-Atlanta, GA 30323 Farmington, CT 06032

l. .INPO Records Center Mr. Darl Hood l Suite 1500 U.S. Nuclear Regulatory Commission l 1100 Circle 75 Parkway Office of Nuclear Reactor Regulation Atlanta, GA 30339 Washington, D.C. 20555 L

i M&M Nuclear Consultants Mr. P.K. Van Doorn L 1221 Avenue of the Americas NRC Resident Inspector l New York, NY 10020 McGuire Nuclear Station l'

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.L.G. Bost l J.S. Warren l R.L. Gill C.L. Hartzell (CNS) i R.S.-Matheson (ONS) ,

P.R. Herran l S.S. Kilborn (W) l R.E. Lopez-Ibanez l M.A'. Mullen 1 R.O. Sharpe (MNS)

G.B. Swindlehurst K.D. "homas M.S. '.'uckman L.E. Feaver R.L. Leber J.D.-Uylie (PSD)

J.W. 'dillis QA Tech. Services NRC Coordinator (EC 12/55)

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LICENSEE EVENT REPORT (LER)

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. . . .,,,e.-. , ,,,, o,,,,,. .u. ,,,..n.,,,, , - n e i On May 12, 1990, at 0229, a Feedwater Isolation occurred on Unit I as a result of Steam Generator IB reaching its Hi-Hi level setpoint of 82% level.

Operationu personnel were in the process of feeding and bleeding the Steam Generators for chemistry control in anticipation of changing from Mode 5 (Cold Shutdown), to Mode 4 (Hot Shutdown) on the following shift. Valve ICF-105, 1B Steam Generator Feed Reg Bypass, was leaking past its seat requiring Operator A to control the level in Steam Generator 1B by opening and closing ICF-127, IB Steam Generator Preheater Bypass Valve. Operator A was assisting in other testing in progress and failed to secure feedwater to the Steam Generator prior to reaching its Hi-Hi level setpoint. Corrective actions included resetting the Feedwater Isolation logic and restoring level to normal.

Operations shift personnel met to discuss the importance of limiting the number of activities requiring control operators attention at any one time.

The computer alarm setpoint for high Steam Generator level will be reduced to provide the Operators an early warning. This event has been assigned a cause of Inappropriate Action resulting from lack of attention to detail.

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EVALUATION:

Background

The Engineered Safety Features Actuation System [EIIS:JF]- (ESFAS) is designed to' actuate ESF equipment in the event predetermined safety limits are exceeded. The analog portion of the ESFAS consists of three to four redundant.

channels of process instrumentation for each parameter or variable that is monitored. The digital. portion of the ESFAS consists of two redundant logic trains which receive inputs from the analog. channels and is a part of the Solid State Protection System (SSPS). . The SSPS performs the logic necessary to determine when and which ESF equipment needs to be actuated, and through systems of relays [EIIS:RLY], actuates the ESF equipment. Technical Specifications do not require t.he ESFAS to be operable in Mode 5.

Each Steam Generator (EIIS:SG] utilizes three level instruments to monitor for

, a high Steam Generator level of M.%,. If two of three of these level instruments see a high level, the SSPS will actuate appropriate. relays to cause the turbine [EIIS:TRB] to trip, 'the feedwater pumps (EIIS:P] to trip, and feedwater to isolate to the Steam Generators.

K Description of Event

,On May 12, 1990 at 0225, Operations personnel were in the process of " feeding and bleeding" the Unit 1 Steam Generators. This involved draining some water from the-Steam Generators by way of the Blowdown System [EIIS:WIl. Drain valves were opened on the Blowdown System which drained water from the Steam Generators to the Turbine Building [EIIS:NM]' Sump. The water was replaced by

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. feeding from the Condensate system [EIIS:KA]. This drain and makeup was being performed in an effort to improve the chemistry of the water in the Steam Generators.

At 0229, the Operator' Aid Computer [EIIS: CPU] alarmed indicating "Hi Steam Generator B Narrow Range Level Channel I". Channels III and IV also alarmed.

At the same time, a Feedwater Isolation signal was received. This signal, caused by the.high Steam Generator level, closed the Steam Generator Preheater Bypass Valves, ICF-126-127-128 and 129, which were being used to feed the Steam Generators. Annunciator [EIIS: ANN] alarm [EIIS: ALM] 1AD-4 D2, Steam

-Generator B Hi-Hi Level _ Alert, alarmed at the same time the isolation signal occurred.

Operator A had been assigned the task of monitoring the Steam Generator levels and adjut, ting Feedwater flow to maintain the levels. This task was complicated by the fact that valve ICF-105, 1B Steam Generator Feed Reg Bypass, was leaking past its seat. This required the operator to periodically open and close valve ICF-127, Steam Generator B Feedwater Preheater Bypass, to maintain the level. Valves ICF-105 and ICF-107 are controlled from the Steam Generator control panel [EIIS:PL).

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0l0 0l3 0 l5 itXT (# sense ansse 4 meusesf, asse essmanaf ##C 7snn MbeW ltM Operator A was also assisting other Operations personnel in the performance of PT/1/A/4350/26, Auxiliary Shutdown Panel Controls Verification.- This procedure requires operations personnel to take control of various pumps and valves at the local Auxiliary Shutdown Panel to ensure these components operate properly from-this location. Operator A was realigning the pumps and valves in the Component Cooling System [EIIS:CC) after the Operators-at the local panel had completed this portion of the procedure when the isolation signal occurred.

Other activities in progress at the time of this event were PT/1/A/4200/28B, Train'B Slave Relay Test. Also, Instrumentation and Electrical (IAE)

Maintenance personnel were working on the Volume Control Tank (EIIS:TKl (VCT)

-level control system under work request 07166A. Another Licensed Control Operator was manually controlling the level in the VCT while this work was in progress.

Unit I was in Mode 5 when the Feedwater Isolation occurred. Station personnel intended to place the Utit in Mode 4 at approximately 0900 on this same morning.

Operations personnel closed the Reactor Trip Breakers at 0305 on May 12, 1990,

,- -after investigating to ensure this would not cause any further problems. The breakers were closed to allow resetting the Feedwater Isolation Signal. The Reactor Trip Breakers were then reopened and the Steam Generator Levels were returned to normal.

The NRC was notified of this event at 0316 as directed by RP/0/A/5700/10, NRC Immediate Notification Requirements.

Conclusion This event is assigned a cause of Inappropriate Action because of lack of-attention to detail. Operator A was assigned the task of monitoring Steam

, fGenerator levels while feeding and bleeding the Steam Generatvc For chemistry control. Operator A was-periodically opening and closing valve ICF-12? to maintain the level in Steam Generator 1B because the control valve, ICF 195, L was leaking past the seat. Operator A was also assisting in the performanc' of the Auxiliary Shutdown Panel Control Verification PT. While assisting on this procedure, he failed to secure feed to Steam Generator 1B prior to teaching the 82% level setpoint for a Feedwater Isolation Signal. -]a, A-mitigating circumstance in this event is the amount of work on going in preparation for changing modes. The Auxiliary Shutdown Panel procedure requires the operator to monitor the response of the various systems being

- tested and their ef fects on the Reactor Coolant System (EIIS: ABl temperature

u. and pressure. This procedure also required the operator to realign portions of the systems being tested. .

Also at the_ time of the event, procedure PT/1/A/4200/28B, Train B Slave Relay Test, was being performed by Performance personnel. This procedure required l

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the control operator to verify equipment conditions before and after segments s of the procedure are. performed. Both control operators had been assisting.

with.this procedure. 4 At the same time, IAE personnel were working on the level control system for (

the VCT. This required one control operator to maintain the VCT level ,

-enually. The second control operator on Unit.1 was performing this task, R is work was being performed as directed by work request 07166A.

mather mitigating circumstance is the avail'able alarms for Steam Generator level. The Steam Generator level deviation alarm is set for +/- 5 percent of program level. In this case, the level deviation alarm came in at 43 percent 7 on the Steam Generator. No other alarms are available to alert the, operators until.the Steam Generator level reaches 82 percent where a Hi-Hi Level Alert annunciator alarms and Feedwater Isolation occurs. A computer alarm for high q Icvel also occurs at this 82 percent level. All four Steam Generators were -[

operating above the + 5 percent deviation alarm because of the feed and bleed  ;

I of the Steam Generators.

This event occurred on the first night of this crew's schedule. All normal ~

-Control Room crew members were present. Operator A han performed feed'and q bleed of the Steam Generators many times in the past. At the time of l, occurrance, Operator A was realigning the component cooling system which is l located approximately fifteen feet from the Steam Generator controls. ,

l t i A. review of the Operating Experience Program database for the past twenty-four l mont.hs prior to this event revealed no events' involving Steam Generator L overfill as a result of Inappropriate Action. Therefore, this event is not l considered recurring.

ESF actuations resulting from Inappropriate Actions is considered a recurring problem.

g This event is not Nuclear Plant Reliability Data System (NPRDS) reportable.

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There were no personnel injuries, radiation overexposures, or uncontrolled releases of radioactive material as a result of this event.

CORRECTIVE ACTIONS:

Immediate: 1) The Feedwater Isolatior. Logic was reset by Operations personnel.

2) Operations personnel restored Steam Generator level to the normal operating range.

Subsequent: 1) Operations personnel performed an NRC notification per RP/0/A/5700/10.

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2) Operations Shift personnel met to discuss this event and to emphasize the importance of limiting the activities in 9 progress to minimize the chance of mistakes.

This event was reviewed with the personnel involved.

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Planned: 1)' The computer alarm setpoint for Hi Steam Generator Narrow Range Level will be reduced-to provide the Operators an early warning.

SAFFTY ANALYSIS:

The purpose of the Feedwater Isolation signal from a Steam Generator Hi-Hi Level is to prevent water from entering the main steam lines. The added weight of the water could potentially challenge the steamline stress limits.

Water in the steam lines could carry over'and damage the Main Turbine or damage,the Auxiliary Feedwater Pump Turbine.

These potential problems were prevented by the closure of the valves supplying 4 water t'o the Steam Generators. Had the valves failsd to close, annunciator

, alarms and computer alarms would have alerted the operators of the situation..

i The operators would have had sufficient time to restore the levels to normal.

Technical Specifications do not require the ESFAS to be operable until Mode 2.

(Startup). The unit was in Mode 5 at the time of occurrance.

The health and safety of the public were not affected by this event.

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