Information Notice 1985-50, Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox

Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox

ML031180244
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill, Crane
Issue date:	07/08/1985
From:	Jordan E NRC/IE
To:
References
IN-85-050, NUDOCS 8507080156
Download: ML031180244 (7)

SSINS No.:

6835 IN 85-50

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C.

20555

July 8, 1985

IE INFORMATION NOTICE NO. 85-50:

COMPLETE LOSS OF MAIN AND AUXILIARY FEEDWATER

AT A PWR DESIGNED BY BABCOCK & WILCOX

ADDRESSEES

All nuclear power facilities holding an operating license (OL) or construction

permit (CP).

Purpose

This information notice is being provided to inform licensees of a significant

reactor operating event involving the loss of main and auxiliary feedwater at

a pressurized water reactor.

Information in this notice is preliminary and was

obtained from the special NRC fact finding team which is investigating the

event.

A complete report of findings will form the basis for further communi- cations or actions related to this event.

The NRC expects that recipients

will review this notice for applicability to their facilities.

Suggestions

contained in this notice do not constitute NRC requirements; therefore, no

specific action or written response is required.

Description of Circumstances

On June 9, 1985, the Davis-Besse plant was operating at 90% power with Main

Feedwater Pump 2 in manual control because problems in automatic had been

experienced.

A control problem with Main Feedwater Pump 1 occurred, and it

tripped on overspeed.

Reactor runback at 50% per minute toward 55% power was

automatically initiated.

Nevertheless, 30 seconds later, the reactor tripped

at 80% power on high pressure in the reactor coolant system.

One second after reactor/turbine trip, one channel of the Steam and Feedwater

Rupture Control System (SFRCS) was automatically initiated due to a spurious

signal indicating low water level in Steam Generator 2. Both Main Steam

Isolation Valves (MSIVs) closed. Three seconds after the actuation, the SFRCS

automatically reset.

Closing of the MSIVs isolated the turbine of the operating

main feedwater pump from its source of steam.

The pump continued to supply

feedwater to the steam generators for a few minutes as it coasted down.

Four and a half minutes after reactor trip, water level in the steam generators

began to fall from the normal post-trip level which is 35 inches.

After MSIV

closure, steam release to atmosphere continued to remove decay heat.

One minute

later, Channel 1 of SFRCS actuated when the water level in Steam Generator 1 actually reached the SFRCS setpoint at 27 inches (See Figure 1).

SFRCS started

Auxiliary Feedwater Pump 1 and initiated alignment of it to Steam Generator 1.

8507080156

IN 85-50

July 8, 1985 Within seconds after automatic initiation of Channel 1 of SFRCS, the operator

actuated both channels of SFRCS; however, he inadvertently actuated both SFRCS

channels on low steam pressure instead of low water level.

When an SFRCS

channel is actuated on low steam pressure, a rupture of the steam line associated

with that channel is presumed to have occurred.

The SFRCS closes the steam

generator isolation valves, including a valve in the auxiliary feedwater line, and aligns the auxiliary feedwater pump to the other steam generator.

Because

both channels had been manually actuated on low steam pressure, both steam

generators were isolated from both auxiliary feedwater pumps.

Five seconds

after the operator's inadvertent actuation of both channels on low steam

pressure, SFRCS Channel 2 received an actual low water level actuation signal.

Because low pressure initiation takes precedence, alignment of the auxiliary

feedwater pumps remained unchanged.

At six minutes into the event as both

auxiliary feedwater pumps were accelerating, they tripped on overspeed.

In summary, all main feedwater had been lost, both steam generators were isolated

from feedwater and were boiling dry, all auxiliary feedwater pumps were tripped, pressure of the reactor coolant system was rising, and reactor coolant system

temperature was increasing.

Within one minute after the operator's inadvertent actuation of the SFRCS on

low steam pressure, the mistake had been recognized and the SFRCS had been

reset.

If equipment had performed in accordance with system design requirements,

-the operator's-error-might not have had-a significant impact on the event.

The auxiliary feedwater isolation valves should have reopened automatically, but the valves did not reopen.

The operator then tried to reopen the valves

from the main control panel, but the valves would not reopen. Operators were

dispatched to locally start the auxiliary feedwater pumps, open the auxiliary

feedwater isolation valves, start the nonsafety-related motor-driven startup

feedwater pump, and valve it to the system.

Pressure and temperature in the reactor coolant system continued to rise

because there was not sufficient water in the steam generators to provide an

adequate heat sink.

At 13 minutes after reactor trip, reactor coolant system

pressure reached 2425 psig, and the Pilot Operated Relief Valve (PORV) opened

three times to limit the pressure rise.

On the third lift, the valve remained

open.

The operator closed the PORV block valve and reopened it two minutes

later after the PORV had closed.

Approximately 16 to 18 minutes after reactor trip, the operators had the startup

and auxiliary feedwater pumps running and the valves aligned. Water levels were

beginning to rise in the steam generators.

Reactor coolant temperature reached

a maximum of 5940 F and then started to decrease to normal.

Refilling of the

steam generators caused the reactor coolant system to fall to 1716 psig and

about 5400F before returning to normal (See Figure 2).

At 30 minutes after reactor trip, plant conditions were essentially stable.

IN 85-50

July 8, 1985 Discussion:

For several minutes after reactor trip, the steam generators were unable to

cool the reactor coolant system adequately.

The first problem contributing to this event was the loss of all main feedwater

due to closure of the MSIVs.

The licensee's hypothesis, based on information

from Babcock & Wilcox, is that turbine trip caused a pressure transient upstream

from the turbine stop valves which caused the outputs of the redundant steam

generator level instrumentation channels to oscillate widely for several

seconds.

The licensee believes that this caused a spurious low level actuation

of SFRCS which closed the MSIVs.

Three additional problems contributed to this event by affecting the availability

of both trains of the auxiliary feedwater system.

The first occurred when the

reactor operator pressed the wrong SFRCS buttons.

The second occurred when

both auxiliary feedwater pumps tripped on overspeed.

The third occurred when

both auxiliary feedwater isolation valves did not reopen when SFRCS was reset.

Control buttons for the SFRCS are arranged in two vertical columns.

Each

column of buttons controls one SFRCS channel.

The operator should have pressed

the fourth button from the top in each column.

Instead, the operator pressed

the top buttons causing isolation of both steam generators.

Both auxiliary feedwater pumps are driven by Terry turbines which tripped on

overspeed early in the event.

When this occurred, steam was being supplied to

the turbines via crossover lines, which are longer than the normal supply lines

and include long horizontal runs.

The licensee believes that significant

condensation may have occurred in the crossover lines.

Further, the licensee

believes that the quality of steam arriving at the turbines may have been

affected significantly by the configuration of the crossover lines and may have

caused the overspeed trips.

The auxiliary feedwater system isolation valves have Limitorque motor operators.

The motor operators have torque switches which prevent overtorquing of the

valves by disconnecting power to the motors. When the valves are being opened, additional torque is required to overcome friction while the gates are being

unseated and while a significant pressure differential may exist across the

gates. During the initial part of the opening stroke, the torque switch in the

motor operator is bypassed by a bypass switch so that full motor torque is

developed if necessary.

The licensee believes that these bypass switches went

off bypass too early.

The valves did not reopen until an operator unseated

them by hand.

IN 85-50

July 8, 1985 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional

Administrator of the appropriate NRC regional office or this office.

ward /Jordan, Director

Division of Emergency Preparedness

and Engineering Response

R. W. Woodruff, IE

(301) 492-4507 Attachments:

1.

Figure 1 - Steam Generator 1 Level and Pressure

2. Figure 2 -

RCS Temperature and Pressure

3.

Attaclnent 1

IN 85-50

July 8, 1985

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IN 85-50

July 8, 1985

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Attachment 3

IN 85-50

July 8, 1985

LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issue

Issued to

85-49

85-48

Relay Calibration Problem

Respirator Users Notice:

Defective Self-Contained

Breathing Apparatus Air

Cylinders

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7/1/85

6/19/85

6/18/85

All power reactor

facilities holding

an OL or CP

All power reactor

facilities holding

an OL or CP, research, and test reactor, fuel cycle and

Priority 1 material

licensees

All power reactor

facilities holding

an OL or CP

85-47

85-46

Clarification Of Several

Aspects Of Removable Radio- active Surface Contamination

Limits For Transport Packages

6/10/85

All power reactor

facilities holding

an OL

85-45

85-44

85-43

85-42 Potential Seismic Interaction 6/6/85 Involving The Movable In-Core

Flux Mapping System Used In

Westinghouse Designed Plants

Emergency Communication

System Monthly Test

Radiography Events At Power

Reactors

Loose Phosphor In Panasonic

800 Series Badge Thermo- luminescent Dosimeter (TLD)

Elements

5/30/85

5/30/85

5/29/85

All power reactor

facilities holding

an OL or CP

All power reactor

facilities holding

an OL

All power reactor

facilities holding

an OL or CP

All power reactor

facilities holding

an OL or CP

OL = Operating License

CP = Construction Permit