Information Notice No. 85-50: Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox
SSINS NO.: 6835
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
July 8, 1985
Information Notice No. 85-50: COMPLETE LOSS OF MAIN AND AUXILIARY
FEEDWATER AT A PWR DESIGNED BY BABCOCK &
All nuclear power facilities holding an operating license (OL) or
construction permit (CP).
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 communications 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 l 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.
July 8, 1985
Page 2 of 4
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
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 594 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 540F before returning to normal (See
At 30 minutes after reactor trip, plant conditions were essentially stable.
July 8, 1985
Page 3 of 4
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.
July 8, 1985
Page 4 of 4
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 region office or this office.
Edward L. Jordan, Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: R. W. Woodruff, IE
1. Figure 1 - Steam Generator 1 Level and Pressure
2. Figure 2 - RCS Temperature and Pressure
3. List of Recently Issued IE Information Notices
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