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 SOR MODEL NUMBER

Penn. Pwr. & Light/Susquehanna 103/B202 So. Cal. Edison/San Onofre 103/B903 TVA/Brown's Ferry 103/8212 TVA/Sequoyah 103/BB212

103/BB203

103/BB803 WPPS 103/BB203 GPU/Oyster Creek 103/B905

103/BB212

103/B212

103/B202 N.E. Nuc./Millstone 103/B903 South Texas Projects 103/BB212

103/BB803 Commonwealth Edison/LaSalle 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- 6/12/86 All power reactor

tamination Of Respiratory facilities holding

Protection Equipment an OL or CP and

fuel fabrication

facilities

86-45 Potential Falsification Of 6/10/86 All power reactor

Test Reports On Flanges facilities holding

Manufactured By Golden Gate an OL or CP and

Forge And Flange, Inc. research and test

facilities

86-44 Failure To Follow Procedures 6/10/86 All power reactor

When Working In High Radiation facilities holding

Areas an OL or CP and

research and test

reactors

86-43 Problems With Silver Zeolite 6/10/86 All power reactor

Sampling Of Airborne Radio- facilities holding

iodine an OL or CP

86-42 Improper Maintenance Of 6/9/86 All power rector

Radiation Monitoring Systems facilities holding

an OL or CP

86-41 Evaluation Of Questionable 6/9/86 All byproduct

Exposure Readings Of Licensee material licensees

Personnel Dosimeters

86-32 Request For Collection Of 6/6/86 All power reactor

Sup. 1 Licensee Radioactivity facilities holding

Measurements Attributed To an OL or CP

The Chernobyl Nuclear Plant

Accident

86-40 Degraded Ability To Isolate 6/5/86 All power reactor

The Reactor Coolant System facilities holding

From Low-Pressure Coolant an OL or CP

Systems in BWRS

OL = Operating License

CP = Construction Permit