Information Notice 1987-60, Depressurization of Reactor Coolant Systems in Pressurized-Water Reactors
UNITED STATES
'NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.
20555
December 4, 1987
NRC 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 result- ing 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 require- ments; 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 PORY
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
M
A0 A
December 4, 1987 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 Mital 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.
December 4, 1987 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 PORYs had been isolated for about 2 weeks prior to the event, also
because of excessive leakage. Although an isolated PORY 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 PORMs 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
December 4, 1987 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
.,
Attachment
December 4, 1987
LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES 1987
"nfornation
Date of
Notice No.
Subject
Issuance
Issued to
86-1089 Supp. 2
87-59
87-58
87-57
87-56
87-55
87-54
87-53
Degradation of Reactor
Coolant System Pressure
Boundary Resulting from
Boric Acid Corrosion
Potential RHR Pump Loss
Continuous Communications
Following Emergency
Notifications
11/19/87
11/17/87
11/16/87
11/6/87
11/4/87
10/29/87
10/23/87
10/20/87
All holders of OLs
or CPs for nuclear
power reactors.
Loss of Emergency
Capability Due to
Gas Intrusion
Foration
All holders of OLs
or CPs for nuclear
power reactors.
All nuclear power
reactor facilities
holding an OL and
the following fuel
facilities that have
Emergency Notification
Systems:
Nuclear
Fuel Services, Erwin, TN; General Atomics, San Diego, CA; UNC,
Montville, CT; and
B & W LRC and B & W
Navy, Lynchburg, VA.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for boiling
water reactors (BWRs).
All NRC licensees
authorized to
possess portable
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
Improper Hydraulic Control
Unit Installation at BWR
Plants.
Portable Moisture/Density
Gauges: Recent Incidents
of Portable Gauges Being
Stolen or Lost
Emergency Response Exercises
Auxiliary Feedwater Pump
Trips Resulting from Low
Suction Pressure
OL = Operating License
CP = Construction Permit
December 4, 1987 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
- SEE PREVIOUS CONCURRENCES
- OGCB: DOEA: NRR
- ROAB:DSP:AEOD
DCKirkpatrick
09/17/87
09/29/87
- PPMB:ARM
TechEd
09/18/87
IN 87-XX
November xx, 1987
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
- SEE PREVIOUS CONCURRENCES
- OGCB:DOEA:NRR
- ROAB:DSP:AEOD
DCKirkpatrick
09/17/87
09/29/87
- PPMB:ARM
TechEd
09/18/87
- C/OGCB:DOEA:NRR
CHBerlinger
10/19/87 D/DOEA:NRR
CERossi
11/ /87
IN 87-XX
September xx, 1987 If the normal spray systems and the PORVs are all unavailable, the auxiliary
spray system can be used to depressurize the primary system.
However, in
this instance, the isolation valve could not be actuated from the control room
because of the loss of a vital bus. Again, the availability of this system is
not controlled by the limiting conditions for operation.
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 regional office or this 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
- SEE PREVIOUS CONCURRENCES
- OGCB:DOEA:NRR
ROAB:DSbAE)p
DCKirkpatrick
SIsrael VLffwY
09/17/87
09/"/87 mq
- PPMB:ARM
TechEd
09/18/87 C/OGCB:DOEA:NRR
CHBerlinger
09/ /87 D/DOEA:NRR
CERossi
09/ /87
IN 87-XX
September xx, 1987
Page of
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 regional office or this 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
OGCB:DOEA:NRR
DCKirkpat~ri4,
09/i7/87MJ *
ROAB:DSP:AEOD
09/ /87 PPMB:ARM C/OGCB: DOEA:NRR
TechE4o CHBerl inger
09//9/87 09/ /87 D/DOEA:NRR
CERossi
09/ /87