Information Notice 1994-90, Transient Resulting in a Reactor Trip and Multiple Safety Injection System Actuations at Salem

Transient Resulting in a Reactor Trip and Multiple Safety Injection System Actuations at Salem

ML031060383
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, 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, Crane
Issue date:	12/30/1994
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-94-090, NUDOCS 9412270233
Download: ML031060383 (11)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555

December 30, 1994 NRC INFORMATION NOTICE 94-90: TRANSIENT RESULTING IN A REACTOR TRIP

AND MULTIPLE SAFETY INJECTION SYSTEM

ACTUATIONS AT SALEM

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

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

notice to alert addressees to the events associated with the loss of

circulating water at Salem Nuclear Power Plant, Unit 1, on April 7, 1994, that

led to a reactor trip followed by multiple automatic actuations of the safety

injection system. 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

On April 7, 1994, at 10:00 a.m., Salem, Unit 1 was in Mode 1 at 73-percent

power. Public Service Electric and Gas Company (the licensee) was operating

the unit at reduced power because river detritus (marsh grass) had fouled the

circulating water intake structure causing a reduction in condenser cooling

efficiency. In response, the operators decreased the power level of Unit 1 to

approximately 60 percent because of an increase in condenser back pressure

caused by grass fouling of the traveling screens at the intake structure. In

response to an impending loss of circulating water, the operators began

reducing load by 1 percent per minute. However, in rapid succession, several

of the Unit 1 traveling screens became clogged with grass, causing the

associated pumps to trip, until only 1 circulating water pump remained

running. As the pumps were lost from service, operators increased the rate of

the load reduction to 8 percent per minute.

Operators attempted to reduce unit load as rapidly as reactor power was being

decreased by insertion of control rods and addition of boron. The effort

caused a power mismatch that resulted in a slight, but continuing, increase in

reactor coolant temperature. In response, the nuclear shift supervisor

directed the operator controlling reactor power to go to the electrical

distribution control panel and shift plant electrical loads to offsite power

sources. Although operators believed that the plant was stable, they failed

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December 30, 1994 to recognize that reactor power was still decreasing because of the delayed

effect of a previous addition of boron. This caused a reversal of the power

mismatch and resulted in reactor coolant system (RCS) temperature decreasing

to below the minimum temperature at which criticality is allowed. The

operators attempted to restore RCS temperature by increasing reactor power

from approximately 7 percent to 25 percent.

However, since power had been

below 10 percent, the power range "high neutron flux-low setpoint" trip had

been automatically reinstated, establishing 25-percent reactor power as the

trip setpoint. When power reached 25 percent, the reactor automatically

tripped.

Almost immediately, train 'A" of the safety injection (SI) logic actuated on a

high steam flow signal coincident with low RCS temperature.

(Later

investigation revealed that the high steam flow signal was actually the result

of a pressure wave created in the main steam lines when the turbine stop

valves closed as a result of the turbine trip).

In response to the reactor

trip and safety injection, the operators entered the plant emergency operating

procedures.

The SI logic did not reposition all necessary components to the

expected, post-actuation position because the initiating signal was so short.

The operators manually repositioned the affected components to their proper

positions. At 11:00 a.m., the licensee declared an unusual event based on a

"manual or automatic emergency core cooling system actuation with a discharge

to the vessel." When the operators took action to reset the SI logic, they

discovered that train "B" of the SI logic had not actuated, indicating an

apparent logic error.

As the operators were attempting to stabilize the plant, the RCS continued to

heat up because of reactor decay heat combined with reactor coolant pump heat.

Steam generator pressure increased but was not automatically relieved by the

steam generator atmospheric relief valves because of a pre-existing condition

that prevented the proper automatic operation of the valves.

Concurrently, because of RCS heatup and the volume of water added by the safety injection, the pressurizer filled to a solid condition, and the pressurizer power- operated relief valves cycled several hundred times to control RCS pressure.

A short time later, steam generator pressure increased in the "11" and "13"

steam generators to the safety valve lift setpoint.

The opening of a safety

valve caused a rapid cooldown and depressurization of the RCS that was

magnified by the solid condition of the system.

RCS pressure rapidly reached

the automatic SI setpoint of 1755 psig, and since train "B" of the SI logic

had remained armed, a second automatic SI actuation occurred. At about the

same time, operators manually initiated safety injection in response to the

rapidly decreasing RCS pressure.

After the second safety injection, operators

remained in the emergency operating procedures, and continued their attempts

to stabilize plant conditions. The pressurizer relief tank rupture disk

actuated because of increasing tank pressure caused by the volume of RCS water

relieved to the pressurizer relief tank from the pressurizer power-operated

relief valves.

IN 94-90

December 30, 1994 The operators controlled plant pressure using the charging and letdown

provisions of the chemical and volume control system because normal RCS

pressure control was not available due to the solid condition of the system.

At 1:16 p.m., licensee management declared an alert to ensure activation of

the Salem Technical Support Center to provide the Salem operators with

additional technical assistance to support cooldown of the plant.

Accordingly, the Technical Support Center was fully staffed.

At 3:11 p.m., the operators established a steam bubble in the pressurizer

using pressurizer heaters.

At 4:30 p.m., operators restored pressurizer level

to the normal band and returned level control to automatic. They subsequently

exited the emergency operating procedures and used the integrated operating

procedures to cool the plant down to Mode 4 (Hot Shutdown), which was achieved

at 1:06 a.m. on April 8, and then to Mode 5 (Cold Shutdown), which was

achieved at 11:24 a.m. on the same day.

Discussion

On April 8, 1994, the NRC dispatched an Augmented Inspection Team to

investigate the event. The results of that inspection were documented in NRC

Inspection Report 50-272/94-80, dated June 24, 1994. Although several issues

emerged from the NRC investigation of this event, three specific aspects are

of particular concern.

These aspects are discussed below.

Solid State Protection System Logic Mismatch: During the first SI actuation, the "A" and "B" logic trains of the solid state protection system were

mismatched.

Train "A" sensed and responded to conditions representative of a

steam line break accident, namely a low RCS temperature coincident with a high

steam line flow.

Although these conditions were real indications, the RCS low

temperature was due to operator error and the high steam flow was a transient

signal induced by a pressure wave resulting from the closure of the turbine

stop and control valves. This transient signal had a duration of about 30

milliseconds, which system response testing later showed was sufficient for

certain portions of the "A" logic to respond, but of insufficient duration for

the "B" logic to respond.

The logic mismatch appears to be a result of the

variations in response sensitivity to the steam flow input relays.

The

licensee modified the design to require a longer signal duration before the

logic is actuated so that such transient signals would not result in an

undesired safety injection.

Nuclear Instrument Rod Shadowing:

Before the initial reactor trip, when the

operators were raising reactor power to restore RCS temperature, the

intermediate range and power range nuclear instruments were not in agreement

with respect to indicated power.

The intermediate range detectors were

"trailing" the power range by about 5 to 10 percent. This led to a condition

in which the reactor was tripped at the 25-percent power range setpoint before

the rod block signal was received from the intermediate range detectors at

20-percent power. The discrepancy between the power and intermediate range

nuclear instruments was apparently due to "rod shadowing."

IN 94-90

December 30, 1994 The combination of the cool RCS and the rod pattern resulting from the down

power maneuver shielded the intermediate range detectors, causing the

instruments to indicate a lower power than the power range detectors.

Although this bias was within an acceptable envelope for detector operability, the response of the instruments was not initially understood.

This led to

concern that the nuclear instruments were not properly operating.

Control Room Command and Control:

Before the initial reactor trip, shift

management directed staff to support actions necessary to restore circulating

water.

The Shift Technical Advisor, a senior reactor operator assigned to the

work control station, was directed to assist in the restoration of affected

equipment. The extra duty reactor operator was directed to assist at the

intake structure.

The senior shift supervisor was initially in the control

room area, but subsequently left to go to the turbine building. This

deployment of licensed operators led to minimal staffing of the control room

at the onset of the transient.

During this time, the operators were preparing to take the unit turbine off

line, and the reactor controls operator was directed by the shift supervisor

to initiate actions to transfer plant electrical loads.

This led to the

reactor controls watch station not being staffed during a reactivity change.

The RCS began to cool as a result of a slight power mismatch between the

reactor and the turbine.

When the shift supervisor first discovered this

mismatch, he began to raise reactor power to restore temperature, which led to

a momentary loss of the command oversight function. He subsequently

recognized the need to maintain an overall command posture and stopped

withdrawing control rods. However, he continued to allow the reactor controls

operator to swap the electrical loads and the RCS temperature continued to

decrease. When the reactor controls operator completed the electrical plant

realignment, the shift supervisor then directed him to raise reactor power to

restore RCS temperature. The shift supervisor did not discuss the fact that

he had manipulated the control rods with the reactor controls operator, and

his direction to the relatively inexperienced operator lacked specificity (how

far or how fast to raise power). The operator subsequently raised reactor

power until the 25-percent power trip was reached.

Related Generic Communications

NRC Information Notice 94-55, "Problems with Copes-Vulcan Pressurizer

Power-Operated Relief Valves," August 4, 1994.

This information notice discusses cracking of plug material, severe wear

of plugs and cages, and a problem with the misalignment and galling of a

stem in the power-operated relief valves discovered as a result of valve

inspection subsequent to the April 7, 1994, event.

NRC Information Notice 94-36, "Undetected Accumulation of Gas in Reactor

Coolant System," May 24, 1994.

This information notice discusses lack of operator awareness of an

accumulation of nitrogen in the reactor vessel head during cooldown and

depressurization of the RCS subsequent to the April 7, 1994, event.

IN 94-90

December 30, 1994

This information notice requires no specific action or written response.

If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

rian K. Grimes, Director

Division of Project Support

Office of Nuclear Reactor Regulation

Technical contacts:

Robert J. Summers, RI

(609) 935-3850

Eric J. Benner, NRR

(301) 504-1171 Attachment:

List of Recently Issued NRC Information Notices

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IN 94-90

December 30, 1994 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issuance

Issued to

94-89

94-88

94-87

94-86

94-85

94-40,

Supp. 1

94-84

89-25, Rev. 1

94-83

Equipment Failures at

Irradiator Facilities

Inservice Inspection

Deficiencies Result in

Severely Degraded Steam

Generator Tubes

Unanticipated Crack in a

Particular Heat of

Alloy 600 Used for

Westinghouse Mechanical

Plugs for Steam Generator

Tubes

Legal Actions Against

Thermal Science, Inc.,

Manufacturer of Thermo-Lag

Problems with the

Latching Mechanism

in Potter and Brumfield

R1O-E3286-2 Relays

Failure of a Rod Control

Cluster Assembly to Fully

Insert Following a Reactor

Trip at Braidwood Unit 2

Air Entrainment in Terry

Turbine Lubricating Oil

System

Unauthorized Transfer of

Ownership or Control of

Licensed Activities

Reactor Trip Followed by

Unexpected Events

12/28/94

12/23/94

12/22/94

12/22/94

12/21/94

12/15/94

12/02/94

12/07/94

12/06/94

All U.S. Nuclear Regulatory

Commission irradiator

licensees.

All holders of OLs or CPs

for pressurized water

reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All fuel cycle and material

licensees.

All holders of OLs or CPs

for nuclear power reactors.

OL = Operating License

CP = Construction Permit

IN 94-XX

November xx, 1994

This information notice requires no specific action or written response.

If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Project Support

Office of Nuclear Reactor Regulation

Technical contacts:

Robert J. Summers

(609) 935-3850

Eric J. Benner

(301) 504-1171 Attachment:

List of Recently Issued NRC Information Notices

E-mailed to John White and Robert Summers of Region I for review on 10/21/94.

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IN 94-XX

November xx, 1994

This information notice requires no specific action or written response.

If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

I

Brian K. Grimes, Director

Division of Projects Support

Office of Nuclear Reactor Regulation

Technical contact:

Robert J. Summers

(609) 935-3850

Eric J. Benner

(301) 504-1171 Attachment:

1.

List of Recently Issued NRC Information Notices

E-mailed to John White and Robert Summers of Region I for review on 10/21/94.

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November xx, 1994

This information notice requires no specific action or written response. If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Projects Support

Office of Nuclear Reactor Regulation

Technical contact:

Eric J. Benner

(301) 504-1171 Attachment:

1.

List of Recently Issued NRC Information Notices

E-mailed to John White and Robert Summers of Region I for review on 10/21/94.

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IN 94-XX

November, xx, 1994 shift supervisor was initially in the control room area, but subsequently left

to go to the turbine building.

This deployment of licensed staffing led to

minimal staffing of the control room at the onset of the transient. During this

time, the operators were preparing to take the unit turbine off-line, and the

reactor controls operator was directed by the shift supervisor to initiate

actions to transfer plant electrical loads. This led to the reactor controls

watch station not being manned during a reactivity change. The RCS began to cool

as a result of a slight power mismatch between the reactor and the turbine. When

first identified by the shift supervisor, he began to raise reactor power to

restore temperature, which led to a momentary loss of the command function. The

shift supervisor subsequently recognized the need to maintain an overall command

posture and stopped withdrawing control rods. However, he continued to allow the

reactor controls operator to swap the electrical loads and RCS temperature

continued to degrade.

When the reactor controls operator completed the

electrical plant realignment, the shift supervisor then directed him to raise

reactor power to restore RCS temperature. This direction was not specific as to

how far to raise power, which, coupled with the operator's inexperience, led to

the operator raising reactor power until reaching the 25 percent power trip.

Related Generic Communications

-

NRC IN 94-55, 'Problems with Copes-Vulcan Pressurizer Power-Operated

Relief Valves,' August 4, 1994.

-

NRC IN 94-36,

Undetected Accumulation of Gas in Reactor Coolant

System," May 24, 1994.

This information notice requires no specific action or written response. If you

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

technical contact listed below or the appropriate Office of Nuclear Reactor

Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Projects Support

Office of Nuclear Reactor Regulation

Technical contact:

Eric J. Benner

(301) 504-1171 Attachments:

1.

List of Recently Issued NRC Information Notices

E-mailed to John White and Robert Summers of Region I for review on l1/19/94.

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IN 94-90

December 30, 1994

This information notice requires no specific action or written response. If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

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Division of Project Support

Office of Nuclear Reactor Regulation

Technical contacts:

Robert J. Summers, RI

(609) 935-3850

Eric J. Benner, NRR

(301) 504-1171 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME:

94-90. IN

• See previous concurrences

E-mailed to John White and Robert Summers of Region I for review on 10/21/94.

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