Information Notice 1986-47, Erratic Behavior of Static O Ring Differential Pressure Switches

Erratic Behavior of Static O Ring Differential Pressure Switches

ML031220689
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	06/10/1986
From:	Jordan E NRC/IE
To:
References
IN-86-047, NUDOCS 8606090487
Download: ML031220689 (6)

SSINS No.: 6835 IN 86-47 UNITED STATES NUCLEAR REGULATORY

COMMISSION

OFFICE OF INSPECTION

AND ENFORCEMENT

WASHINGTON, DC 20555 June 10, 1986 IE INFORMATION

NOTICE NO. 86-47: ERRATIC BEHAVIOR OF STATIC "O" RING DIFFERENTIAL

PRESSURE SWITCHES

Addressees

All boiling water reactor (BWR) and pressurized

water reactor (PWR) facilities

holding an operating

license (OL) or a construction

permit (CP).

Purpose

This information

notice is intended to advise licensees

of erratic behavior of certain differential

pressure switches supplied by SOR, Incorporated (formerly Static "O" Ring Pressure Switch Company) which apparently

caused failure of the LaSalle 2 reactor to scram automatically

when it was operating

with water level below the low level setpoint.

Similar switches are also installed

in the high pressure core spray system and the residual heat removal system.It is expected that recipients

will review this information

for applicability

to their reactor facilities

and consider actions, if appropriate, to preclude the occurrence

of a similar problem at their facility.

Suggestiohs

contained in this notice do not constitute

NRC requirements.

Therefore, no specific action or written response is required.The NRC evaluation

of this incident is continuing.

If specific action is determined

to be necessary, a separate notification

will be issued.Summary of Circumstances

On June 1, 1986, LaSalle 2 experienced

a feedwater

transient

that resulted in a low reactor water level. One of the four low level trip channels actuated, resulting

in a half scram. The operator recovered

level and operation

was continued.

Subsequent

reviews by licensee personnel

raised concerns that the level had apparently

gone below the scram setpoint and thus a malfunction

of the reactor scram system may have occurred.

Based on this concern, the licensee declared an "Alert" and shut the plant down. The NRC dispatched

an augmented inspection

team to the site. Subsequently, the licensee found that the "blind" switches which operate on differential

pressure perform erratically.

The licensee also found erratic operation

for similar switches in the high pressure core spray system and the residual heat removal system which operate valves in the minimum flow recirculation

lines. Based on these results, the licensee declared all emergency

core cooling systems in LaSalle 1 and 2 to be inoperable.

Both units are in cold shutdown pending further evaluation

of the problem.8606090487 IN 86-47 June 10, 1986 Description

of Circumstances:

The following

description

was constructed

from a preliminary

sequence of events prepared by the augmented

inspection

team and from other input by the team.At 4:20 A.M. on Sunday, June 1, 1986, LaSalle 2 was operating

at 93 percent of full power. Both turbine-driven

feedwater

pumps were operating, with the "A" pump in manual control and the "B" pump in automatic

control. The motor-driven

feedwater

pump was in standby. While a surveillance

test was being conducted on feedwater

pump "A", the turbine governor valve opened further and caused pump speed and reactor water level to start increasing.

At about the same time, the automatic

control systems for both turbine-driven

pumps locked out. The reactor operator regained control of feedwater

pump "A" and ranback feedwater

pump speed in an attempt to restore water level to the nominal value (36 inches on the narrow range recorder).

A few seconds later when the control system was reset, the "B" feedwater

pump controller

automatically

ranback the pump speed to zero for no apparent reason. Reactor water level started falling at about 2 inches/second.

Subsequently, the reactor protection

system responded

via separate level switches to the falling reactor water level by reducing recirculation

flow to reduce power, and the operator started the motor-driven

feedwater

pump to increase level. The level continued

to fall for a few more seconds before turning around. The minimum reactor scram setpoint required in the technical

specification

is 11 inches. The level channels are normally set to trip at 13.5 inches, and the operators

are trained to expect reactor scram by the time that the water level reaches 12.5 inches. As the level was falling, one of the four reactor scram level switches (the "0" switch) tripped at approximately

10 inches, causing a"half scram." As designed, this did not initiate control rod motion. None of the other three level switches tripped during this transient.

No reactor scram occurred during this transient, either automatically

or manually.In the BWR scram system logic, which is one-out-of-two-taken-twice, at least one instrument

channel in each scram system must trip to generate a scram demand signal and thereby initiate control rod motion. Preliminary

results of the investigation

indicate that the reactor water level fell to a minimum value of about 4.5 inches on the narrow range instrumentation, which is several inches below the specified

scram setpoint but still 13 to 14 feet above the top of reactor fuel. The period that the water level was below the specified scram setpoint value was approximately

2 seconds. After feedwater

flow turned the transient

around, the plant stabilized

at a power level of about 45 percent.The "B" scram system half scram was manually reset about 30 seconds later. The power level was increased

to 60 percent about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> later.Shortly after the subsequent

shift change, the oncoming shift engineer's

review was effective

in indicating

that the reactor water level appeared to have fallen below the scram setpoint and the level switches may not have performed

properly.He then requested

that an instrumentation

technician

check the calibration

of the switches.

The results were that the "A" and "C" switches, which are in the"A" scram system, tripped at 10 and 13.5 inches respectively

during the calibration

check; the "B" and "D" switches, which are in the "B" scram system, tripped at 11 and 13.5 inches respectively.

The switches were readjusted

to

IN 86-47 June 10, 1986 trip at 13.5 inches. Based on these results, the operating

staff believed that a malfunction

of the scram system may have occurred.

An orderly shutdown of the plant was initiated

at 2:00 P.M. (COT). At 2:30 P.M., the resident inspector was notified, and at 5:30 P.M., the NRC Operations

Center was called via the emergency

notification

system and informed of this event by the licensee.At 6:20 P.M., the licensee decided that the "A" scram system had failed to perform during the transient.

The "A" scram system was manually tripped providing

a half scram on the side that had apparently

malfunctioned.

The orderly shutdown was continued, and an "Alert" was declared.

When all the control rods had been fully inserted at 9:22 the next morning, the Alert was terminated.

On Monday, June 2, the NRC determined

that the incident warranted

a thorough investigation.

The NRC Regional Administrator

dispatched

an augmented

inspection

team to the plant site.On Monday evening, June 2, the licensee checked the calibration

of the reactor scram water level switches by varying the actual level in the vessel. The results were that the "A" and "C" switches tripped at indicated

levels of 9.0 and 6.9 inches respectively

and the "B" and "D" switches tripped at 3.9 and 10.2 inches respectively.

These data were obtained about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the switches had been calibrated

according

to plant procedures

and suggest a non-trivial

difference.

Additional

data obtained over the next two days by varying reactor water level demonstrated

continued

erratic behavior of switch setpoints.

On Saturday, June 7, after calibrating

the Static "O" Ring flow switch which actuates the minimum flow recirculation

valve in the high pressure core spray system, the licensee performed

a different

test using actual system flow. The switch actuated when flow was at 530 gpm instead of 1000 gpm where it had been set to actuate. The licensee found similar performance

of flow switches in the residual heat removal system. The licensee now suspects all Static "O" Ring differential

pressure switches and has declared all emergency

core cooling systems in both units to be inoperable.

Both units remain in cold shutdown.Discussion:

It appears at present that the water level decreased

below the scram setpoint for about two seconds and reached a minimum level of about 4.5 inches. This is based on a recording

from the narrow range water level instrument

and records from the startup testing data acquisition

system which recorded levels from the same transmitter.

Had the reactor operator been aware of this fact before the water level had increased

to a level above the setpoint, tie reactor operator would have been expected to scram the reactor manually.The differential

pressure switches which provide the water level trip input to the reactor scram system were provided by SOR, Incorporated.

These level switches are not original equipment;

but were installed

during replacement

of equipment in secondary

containment.

Affected licensees

had determined

that the original switches were not qualified

to operate in the environment

created by an accident.Operation

of the SOR switches has been demonstrated

to be erratic with little correlation

between the setpoints

established

during atmospheric

pressure

IN 86-47 June 10, 1986 calibrations

and switch actuations

under system pressure conditions.

Exercising

the switches by applying successive

differential

pressure cycles appears to mask erratic setpoint behavior.

Similar problems with SOR differential

pressure switches have been reported at Oyster Creek.Per plant procedure, the switches for reactor water level had been exercised prior to calibration

following

failure of the reactor to scram automatically.

For this reason, performance

of the level switches may have been different

during calibration

than during the event. Further, none of the level switches in the LaSalle 2 reactor scram system operate in conjunction

with individual

level transmitters.

Therefore, the calibration

and performance

of the individual

low level trip channels cannot easily be compared to each other. In effect, the operator is blind to switch performance.

The vendor has indicated

that those plants identified

in Attachment

1 have similar differential

pressure switches.

This list of plants includes pressurized

water reactors as well as boiling water reactors.

NRC intends to meet with representatives

of General Electric Company, SOR Incorporated, and interested

licensees

at 10 A.M. on Thursday, June 12, 1986, in Bethesda, Maryland to discuss experience

with the switches.It is suggested

that licensees

consider advising their reactor operators

of the LaSalle incident and providing

guidance to them as to how to promptly detect the occurrence

of a similar problem at their plants and the proper remedial action to be taken.No specific action or written response is required by this notice. If you have any questions

regarding

this matter, please contact the Regional Administrator

of the appropriate

regional office or this office.or an Divisi of Emergency

Preparedness

and Egineering

Response Office of Inspection

and Enforcement

Technical

Contacts:

J. T. Beard, NRR (301) 492-4415 Roger W. Woodruff, IE (301) 492-7207 Attachments:

1. Plants with Similar Differential

Pressure Switches 2. List of Recently Issued IE Information

Notices

Attachment

1 IN 86-47 June 10, 1986 PLANTS WITH SIMILAR DIFFERENTIAL

PRESSURE SWITCHES PLANT Penn. Pwr. & Light/Susquehanna

So. Cal. Edison/San

Onofre TVA/Brown's

Ferry TVA/Sequoyah

WPPS GPU/Oyster

Creek N.E. Nuc./Millstone

South Texas Projects Commonwealth

Edison/LaSalle

SOR MODEL NUMBER 103/B202 103/B903 103/8212 103/BB212 103/BB203 103/BB803 103/BB203 103/B905 103/BB212 103/B212 103/B202 103/B903 103/BB212 103/BB803 103/B202 103/8212 103/B203 103/BB203 103/BB212 103/BB205 103/68202 Attachment

2 IN 86-47 June 10, 1986 LIST OF RECENTLY ISSUED IE INFORMATION

NOTICES Information

Date of Notice No. Subject Issue Issued to 86-46 Improper Cleaning And Decon-tamination

Of Respiratory

Protection

Equipment Potential

Falsification

Of Test Reports On Flanges Manufactured

By Golden Gate Forge And Flange, Inc.6/12/86 6/10/86 86-45 86-44 Failure To Follow Procedures

6/10/86 When Working In High Radiation Areas 86-43 Problems With Silver Zeolite Sampling Of Airborne Radio-iodine 86-42 86-41 Improper Maintenance

Radiation

Monitoring

Of Systems 86-32 Sup. 1 86-40 Evaluation

Of Questionable

Exposure Readings Of Licensee Personnel

Dosimeters

Request For Collection

Of Licensee Radioactivity

Measurements

Attributed

To The Chernobyl

Nuclear Plant Accident Degraded Ability To Isolate The Reactor Coolant System From Low-Pressure

Coolant Systems in BWRS 6/10/86 6/9/86 6/9/86 6/6/86 6/5/86 All power reactor facilities

holding an OL or CP and fuel fabrication

facilities

All power reactor facilities

holding an OL or CP and research and test facilities

All power reactor facilities

holding an OL or CP and research and test reactors All power reactor facilities

holding an OL or CP All power rector facilities

holding an OL or CP All byproduct material licensees All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP OL = Operating

License CP = Construction

Permit