Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
IN 87-60
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
December 4, 1987
Information Notice No. 87-60: DEPRESSURIZATION OF REACTOR COOLANT
SYSTEMS IN PRESSURIZED-WATER REACTORS
Addressees:
All holders of operating licenses or construction permits for pressurized
water reactors.
Purpose:
This notice is being provided to alert addressees of potential problems
resulting from the loss of pressure control in the reactor coolant system
(RCS) which could affect the operator's ability to depressurize the reactor
coolant system in a timely manner during a steam generator tube rupture
accident, or to control the plant during natural circulation cooldown. 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 do not
constitute NRC requirements; therefore, no specific action or written response
is required.
Description of Circumstances:
Two events have occurred which demonstrate the importance of maintaining the
capability to depressurize the RCS in emergencies.
The importance of maintaining effective pressure control in mitigating a steam
generator tube rupture event was positively demonstrated during the North Anna
Unit 1 tube rupture which occurred on July 15, 1987. A double ended rupture
of a single tube occurred in the "C" steam generator causing an initial break
flow of around 600 gpm.
The plant was manually tripped from 100% power at about five minutes into the
event. This was followed in about 20 seconds by an automatic safety injection
actuation. After positively identifying and isolating the steam generator
with the rupture, the operators initiated a rapid cooldown to 480 degrees F in
order to establish an adequate subcooling margin. This was accomplished by
dumping steam from the undamaged steam generators to the main condenser using
steam dump valves. A few minutes later a rapid RCS depressurization was
commenced by fully opening the two pressurizer spray valves. As this pressure
reduction began to tail off, the operators briefly opened a pressurizer PORV
causing an additional rapid 40 psi drop in RCS pressure. The primary to
secondary leakage was promptly terminated when the RCS pressure was equalized
with the pressure of the steam generator having the ruptured tube at about 30
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minutes into the event. During the remainder of the cooldown the primary
pressure was maintained below the pressure of the steam generator with the
rupture to minimize secondary contamination and to facilitate cooldown of the
steam generator using backfill.
Because they were able to maintain good primary pressure control, the
operators were able to minimize the radiological release during this event.
None of the secondary atmospheric relief valves were actuated. The release
was limited to the contamination of the secondary system before the steam
generator with the rupture could be isolated. The total release was estimated
at 159 mCi for the entire event.
On August 26, 1986, a reactor trip occurred at Salem Unit 2 when a technician
inadvertently grounded a 120 VAC instrument bus, causing a spurious loss-of-
reactor-coolant-pump reactor trip signal. The voltage spike generated by the
grounding also generated a spurious low-steam line pressure signal which, in
conjunction with a high-steam flow indication, initiated a safety injection
signal. About 30 seconds later, a series of vital bus transfers were
generated by the protective relaying logic. During these transfers, two of
the vital buses were without power simultaneously for about two seconds, which
resulted in the generation of a station blackout signal. However, offsite
power was actually available and the reactor coolant pumps continued to
operate. The coincident safety injection and station blackout signals
disconnected all vital power buses and automatically sequenced selected safety
injection loads onto the emergency buses powered by the already operating
diesel generators. The number two vital bus remained deenergized because the
diesel generator for this bus had been taken out of service for maintenance.
However, in accordance with the plant design, this automatic sequencing did
not load the component cooling water pumps onto the emergency buses.
The reactor operators secured the reactor coolant pumps after 5 minutes of
operation because component cooling water was not available to cool the motor
bearings and the thermal barrier. The high-pressure safety injection pumps
continued to operate after the receipt of the safety injection signal. The
resulting rise in reactor coolant system pressure caused a power-operated
relief valve (PORV) to lift numerous times. Normal pressurizer spray was not
available to control the primary system pressure rise once the reactor coolant
pumps were tripped.
Although safety injection was not needed, the charging pumps continued to
inject water into the primary system through the emergency core cooling system
(ECCS) piping. The isolation valves had assumed their safeguards (open)
position following initiation of the safety injection signal. Since the vital
bus that powered the ECCS isolation valves was deenergized, the control room
operators could not isolate the ECCS flow without taking local manual control
of the isolation valves. The operators elected not to shutdown the charging
pumps because they were needed to supply injection water to the reactor
coolant pump seals. In addition, the operators were unable to initiate the
auxiliary pressurizer spray even with the charging pumps running because the
spray isolation valve, also connected to the deenergized vital bus, was closed
as part of the automatic safeguards alignment and could not be opened
remotely.
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The operators manually energized the component cooling water pumps after 7
minutes. However, it took more than 20 minutes for the operators to secure
safety injection, start a reactor coolant pump, and reestablish normal
pressure control.
Discussion:
Reactor coolant system pressure control is necessary for the timely recovery
from steam generator tube rupture accidents; i.e., to minimize the discharge
of reactor coolant into the faulted steam generator and the subsequent loss of
coolant outside containment, such as occurred during the Ginna accident
(January 25, 1982). Pressure control also is important to facilitate natural
circulation cooldown. Generally, the normal pressurizer spray system is used
to control or reduce reactor coolant system pressure. However, this system
requires the operability of the reactor coolant pumps and the pressurizer
spray control valves. In the Salem event, the reactor coolant pumps were
secured and one of the normal pressurizer spray lines had been isolated for
about three months because of excessive leakage.
Emergency operating procedures for many plants utilize the PORVs for depres-
surizing the primary system following a steam generator tube rupture accident
if the normal pressurizer spray system is not available. In the Salem event,
one of the PORVs had been isolated for about 2 weeks prior to the event, also
because of excessive leakage. Although an isolated PORV could probably be
unblocked if it was seriously needed for pressure reduction, the PORV
isolation represents an additional loss of redundancy and reliability. If the
normal pressurizer spray system is out of service and the PORVs are
unavailable, the auxiliary pressurizer spray system on plants having such a
system can be used to depressurize the primary system. However, during the
Salem event the auxiliary pressurizer spray system was also unavailable
because its isolation valve was closed and could not be repositioned from the
control room due to the loss of its vital bus. This vital bus was not
re-energized immediately because the diesel generator supplying power to this
bus was out of service for preventive maintenance.
The availability of the pressurizer spray system, the PORVs for some plants
and the auxiliary pressurizer spray system are generally not assured by the
limiting conditions for operation contained in the Technical Specifications.
Nevertheless, as these events demonstrate, these systems can be important to
the safety of the plant under certain emergency conditions. Consequently, it
is important that out of service periods for repairs or maintenance be
minimized for these systems. In the case of the PORVs the reliability of the
closing capability as well as the assurance of availability for pressure
control is important. During the Ginna accident the PORV stuck open causing a
loss-of-coolant to the containment and the formation of coolant voids in the
reactor vessel head and the tube bundle of the faulted steam generator. At
Indian Point Unit 2 (LER 247/85-002) and Callaway Unit 1 (LER 483/84-064) all
of the PORVs were found to have been isolated during normal operation,
inhibiting their ability to provide pressure control and to promptly mitigate
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a potential accident. Further information regarding this issue can be found
in AEOD/E708, "Depressurization of Reactor Coolant Systems in PWRs," an engi-
neering evaluation report issued by the NRC Office for the Analysis and
Evaluation of Operational Data.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the technical
contact listed below or the Regional Administrator of the appropriate regional
office.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Sanford Israel, AEOD
(301) 492-4437
Donald C. Kirkpatrick, NRR
(301) 492-8166
Attachment: List of Recently Issued NRC Information Notices
