Information Notice 1992-49, Recent Loss or Severe Degradation of Service Water Systems

Recent Loss or Severe Degradation of Service Water Systems

ML031200193
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	07/02/1992
From:	Rossi C Office of Nuclear Reactor Regulation
To:
References
IN-92-049, NUDOCS 9206290235
Download: ML031200193 (5)

I C. AL

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 July 2, 1992 NRC INFORMATION NOTICE 92-49: RECENT LOSS OR SEVERE DEGRADATION OF

SERVICE WATER SYSTEMS

Addressees

Is

.I

All holders of operating licenses or construction ormits for nuclear power

reactors.

'-.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is isating this information

notice to alert addressees to recent operating exp5dience problems involving

the loss or potential loss of safety-related heat transfer capability in

service water systems. -It is expected that recipients will review the

information for applicability to their-facilities and'consider actions, as

appropriate, to avoid similar problems. However, suggestions contained in

this information notice are not NRC requirements; therefore,'-no specific

action or written response is required.

Description of Circumstances

Nine Mile Point. Unit 1. February 21. 1992. When.performing post-maintenance

testing while the' reactor was shut down, the licensee,'Niagara Mohawk Power

Corporation, inadvertently isolated the ultimate heat sink by closing all

gates to the SWS inlet bay. Because one SWS pump and two circulating water

pumps were running, the water level in the bay.rapidly decreased. For about

6 minutes, the level was below that, assumed in the licensing basis and below

the minimum level necessary to maintain net positive suction head for any of

the SWS pumps'in the'bay. The running SWS pump cavitated; the licensee

started the emergency SWS pump as required by procedures, but then had to stop

it because of low discharge pressure (NRC Augmented Inspection Team . -

(AIT) Report 50-220/92-80).

The licensee had aligned the gates-innthe intake SWSbay in an off-normal.

configuration for reverse'flow to allow post-maintenance testing of the gate D

opening circuit.. The licensee uses the reverse flow configuration to prevent

icing during winter-months. The maintenance included the removal of an

undocumented electrical jumper used to bypass'the mechanical tension overload

protection switch in the drive motor circuit. 'The licensee did not know if

the gate could be opened or closed during reverse flow operation with the

jumper removed.' After closing the~gate,' the licensee'could not then reopen

it. 'Within'-2-3 minutes'the level decreased t6,a point-'where neither the_

normal SWS pumps nor the emergency SWS pumps could maintain adequate suction.

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IN 92-49

-July 2, 1992 - The licensee promptly opened gate D by jumpering the tension overload switch

and also opened one of the normal lineup inlet gates. The gates take about

5 minutes to fully open. The intake bay level returned to normal within

6 minutes and both emergency SWS pumps were successfully started within

another 3 minutes.

The AIT concluded that the root causes of this event were: the failure to

follow the established work control process, ,inadequate management oversight, inadequate communications within and among organizations participating in the

work activities, and an insensitivity to shutdown risk among multiple licensee

organizations.

Arkansas Nuclear One. Unit 2. April 16. 1991. The licensee, Arkansas.Power

and Light Company,, declared both loops of the safety-related SWS inoperable

with the reactor in startup conditions. Debris from thejlake, the normal

supply of cooling water, had bypassed the screens-at the pump suctions and

clogged the pump discharge strainers of both operating loops. Fortunately, the standby SWS pump was not operating at the time and its discharge strainer

remained free of debris. The licensee switched'tfie'.suction of the standby

SWS pump to its emergency source and started the pump within about 3 minutes.

The licensee restored the clogged loops to operable status within about

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (LER 50-368/91-12).

The loss of both SWS trains meant that cooling was not available to engineered --

safety features equipment and component cooling water heat exchangers, which

cool systems carrying fluids that may be radioactive. This condition resulted

when maintenance personnel performed sections of the procedure for rotating

the traveling screens-out-of sequence and; thus, allowed screen rotation

without wash water flow. Consequently, the flow of debris bypassed the

traveling screens and entered the suction of the two operating SW pumps. Had

the standby SW pump discharge strainer become clogged, the event would have

been much more-severe.

Ineffective communication between operations and maintenance personnel

prevented a complete understanding of the operation in progress at the time.

The licensee took steps to strengthen management control and the training of

personnel in this procedure and in communications.

Fitzpatrick, Oct6ber 19. 1990. The licensee, Power Authority of'the State of

New York, manually scrammed the reactor from 45 percent of full power because

the fouling rate-for the circulating water traveling screens exceeded the

cleaning rate of the screen wash system. A shift in wind direction

contributed to an unusually large debris accumulation'on the'screens.- Shear

pins on-the two operating screens failed. As the screens bowed inward because

of the high differential pressure, some of the debris floated around the

screens. The-licensee scrammed the reactor to mitigate this degrading

condition (LER 50-333/90-23).

In this event, while~performing maifitdnan'ce on one of the traveling screens,

<Ad pe'rsonnel'unintentionally'disabled the screen differential pressure alarm

system, which would have provided early indication of fouling. The licensee

determined the root cause of the event to be that the applicable operating and

_j

IN 92-49 July 2, 1992 maintenance procedures did not specify the need to isolate the differential

pressure instrument system from only the specific intake bay that is drained.

Millstone Unit 1. October 4. 1990. The licensee, Northeast Nuclear Energy

Company, manually tripped the reactor from 45 percent of full power because of

circulating water system and service water system fouling that resulted in

degraded SWS cooling, which resulted in increased containment temperature and

pressure. Storm-induced high winds and seas caused an excessive amount of

seaweed to accumulate on the traveling screens of the circulating water

system. After first questioning the off scale indication of the differential

pressure instruments, the licensee stopped two of the four operating

circulating water pumps to relieve stress on the screens, but the relief was

insufficient to prevent three of the five screens from collapsing. The two

operating pumps reduced the water level in the intake structure bays, which

caused the operating SWS pumps to cavitate. This condition decreased

SWS pressure, degraded the performance of the reactor building closed cooling

water heat exchanger, increased the containment temperature and pressure, and

decreased the main condenser vacuum (LER 50-245/90-16).

The SWS provides cooling to the turbine and the reactor building closed

cooling water heat exchangers and to the heat exchangers for the diesel

generators. The emergency SWS provides long-term cooling to the suppression

pool during a loss-of-coolant accident (LOCA). The licensee noted that the

concurrent loss of these systems with a LOCA is outside the design basis for

Millstone Unit 1.

The licensee concluded that if all the circulating water pumps had been

tripped on increasing differential pressure, the three damaged screens might

not have been breached, and SWS performance might not have been degraded. The

licensee delayed its decision to trip circulating water pumps because control

room personnel had not been informed that plant equipment operations personnel

had disabled all the screens for manual cleaning. The control room personnel

did not trip all circulating water pumps as required by the applicable

operating procedure.

Discussion

"Operating Experience Feedback Report - Service Water System Failures and

Degradations in Light Water Reactors," NUREG-1275, Volume 3, November 1988, summarized and discussed service water system (SWS) events from 1980 to

early 1987. Generic Letter 89-13, "Service Water System Problems Affecting

Safety-Related Equipment," July 18, 1989, requested specific licensee actions

to resolve SWS problems.

The Nine Mile Point Unit 1 event shows that personnel errors and failure to

follow procedures can cause the safety-related SWS to become inoperable.

The other 3 events are examples in which intake debris, caused by adverse

environmental conditions, together with personnel errors, either caused or

could have caused the safety-related SW system to become inoperable. All four

events illustrate that recovery strongly depends on human action, particularly

with respect to following procedures and accurately communicating information.

IN 92-49 July 2, 1992 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

h~arles E. i, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical contacts: John Thompson, AEOD

(301) 492-8091 Vern Hodge, NRR

(301) 504-1861 James Tatum, NRR

(301) 504-2805 Attachment: List of Recently Issued NRC Information Notices

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Attachment

IN 92-49 July 2, 1992 Page 1 of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

92-48 Failure of Exide Batteries 07/02/92 All holders of OLs or CPs

for nuclear power reactors.

92-47

Intent

ional Bypassing 06/29/92 All holders of OLs or CPs

of Automatic Actuation for nuclear power reactors.

of Plant Protective

Features

92-46 Thermo-Lag Fire Barrier 06/23/92 All holders of OLs or CPs

Material Special Review for nuclear power reactors.

Team Final Report Findings, Current Fire Endurance

Tests, and Ampacity Cal- culation Errors

92-45 Incorrect Relay Used in 06/22/92 All holders of OLs or CPs

Emergency Diesel Generator for nuclear power reactors.

Output Breaker Control

Circuitry

92-44 Problems with Westing- 06/18/92 All holders of OLs or CPs

house DS-206 and DSL-206 for nuclear power reactors.

Type Circuit Breakers

92-43 Defective Molded Phen- 06/09/92 All holders of OLs or CPs

olic Armature Carriers for nuclear power reactors.

Found on Elmwood Con- tactors

92-42 Fraudulent Bolts in 06/01/92 All holders of OLs or CPs

Seismically Designed for nuclear power reactors.

Walls

92-41 Consideration of the 05/29/92 All holders of OLs or CPs

Stem Rejection Load in for nuclear power reactors.

Calculation of Required

Valve Thrust

OL = Operating License

2P = Construction Permit