Information Notice No. 94-12: Insights Gained From Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
February 9, 1994
NRC INFORMATION NOTICE 94-12: INSIGHTS GAINED FROM RESOLVING GENERIC
ISSUE 57: EFFECTS OF FIRE PROTECTION SYSTEM
ACTUATION ON SAFETY-RELATED EQUIPMENT
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the insights the NRC staff gained from resolving
Generic Issue (GI) 57, "Effects of Fire Protection System Actuation on Safety-
Related Equipment." 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 are not NRC requirements; therefore, no specific
action or written response is required.
Description of Circumstances
The resolution of GI-57 involved gaining a detailed understanding of the
potential safety significance of fire protection system intended and
inadvertent actuations at U.S. commercial nuclear power plants. During the
resolution process, the NRC staff reviewed operational experiences involving
fire protection system actuations and developed a methodology for quantifying
the effects of such actuations on safety-related equipment. The staff applied
this methodology to one boiling-water reactor (BWR) and three pressurized-
water reactors (PWRs). In doing this, the staff conducted extensive plant
walkdowns and detailed reviews of plant documentation. Building on the
insights gained from the analysis of these four plants, the staff also
performed a generic risk assessment.
Discussion
The insights presented in this information notice stem from the experience
base developed from the detailed study of four operating light-water reactor
designs (References 1 - 4), as well as from a generic risk assessment
developed in Reference 5 which is summarized in the regulatory analysis for
resolving this issue (Reference 6). Attachment 1 summarizes the more
significant insights developed during the study. Attachment 2 lists the
references.
9402030011.
IN 94-12
February 9, 1994
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The risk reduction estimates, cost/benefit analyses, and other insights gained
from resolving GI-57 show that consideration of the matters contained in this
information notice (details are given in Reference 6) can reduce risk due to
fire protection system actuations. However, in view of the observed large
differences in plant-specific characteristics associated with the effects of
such actuations, plant-specific analyses would be required to identify risk
reductions. Plant-specific analyses of the type needed for this purpose are
being carried out as part of the Individual Plant Examination of External
Events (IPEEE) program, recommended by Generic Letter 88-20, Supplement 4,
issued June 28, 1991.
Related Generic Communications
Information Notice 83-41, "Actuation of Fire Suppression System Causing
Inoperability of Safety-Related Equipment"
Information Notice 85-85, "Systems Interaction Event Resulting in Reactor
System Safety Relief Valve Opening Following a Fire-Protection Deluge System
Malfunction"
Information Notice 86-106, Supplement 2, "Feedwater Line Break"
Information Notice 87-14, "Actuation of Fire Suppression System Causing
Inoperability of Safety-Related Ventilation Equipment"
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
/s/'d by BKGrimes
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: H. W. Woods, RES
(301) 492-3908
P. M. Madden, NRR
(301) 504-2854
Attachments:
1. Summary of the Most Significant Insights
Concerning the Effects of Fire Protection
System Actuation on Safety-Related Systems
2. References
3. List of Recently Issued NRC Information Notices.
Attachment 1
IN 94-12
February 9, 1994
Page 1 of 3
SUMMARY OF THE MOST SIGNIFICANT INSIGHTS CONCERNING THE EFFECTS
OF FIRE PROTECTION SYSTEM ACTUATION ON SAFETY-RELATED SYSTEMS
The six most significant insights gained by the NRC staff during the study of
the effects of fire protection system (FPS) actuation on safety-related
equipment are:
1. Mercury Relays
a. Mercury relays were present in the fire protection control systems for
a diesel generator (DG) room. These relays are susceptible to seismic
actuation. If present in common with any of the following features
(identified on other plants), the potential for station blackout
during a seismic event is increased:
1) Water deluge-type FPSs in the DG rooms with nozzles aimed at the
DG control panel, diesel air intake, or generator cooling air
intake.
2) Fire protection control systems that lock out the diesel
generators and/or isolate the diesel generator rooms' cooling when
the FPS is actuated in the DG rooms.
3) A CO2 FPS in a DG room where the DG control system is designed to
shut down the engine due to presence of high CO2 or low oxygen in
the engine air intake.
b. Mercury relays were present in an auxiliary FPS control circuit
designed to isolate cooling in a high-pressure coolant injection
(HPCI) pump room. This design could result in the loss of the HPCI
pump as the room overheats following a seismic event.
c. Mercury relays were present in the actuation circuits for a control
room Halon FPS. An inadvertent release of Halon could require either
donning of emergency breathing apparatus (thus compounding
communications problems and increasing the probability of human
errors) or abandoning the control room following a seismic event.
2. Seismic Dust/Smoke Detectors
Smoke detectors present in the fire protection actuation systems in many
plants will likely be actuated by the dust that rises during a seismic
event. When a fire protection control system is actuated by smoke
detectors alone, a seismic event has the potential to lead to an
inadvertent release of suppressant. A design of this type was observed
for the CO2 FPS in a cable spreading room.
.
Attachment 1
IN 94-12
February 9, 1994
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3. Water Deluge Systems
Critical cabinets with open conduit penetrations on top, or any non-
sprayproofed, safety-related cabinets or components that can be sprayed by
deluge system spray heads are susceptible to damage. For example, the
control panel, the diesel engine air intake, and the electric generator
cooling air intake on DG units are vulnerable if water deluge nozzles are
aimed to spray on any of these areas.
4. Fire Suppressant Availability During a Seismic Event
a. One water FPS was installed with one pump driven by an electric motor
and the other driven by a diesel engine. During a seismic-related
loss of offsite power, the electric pump's non-vital power source
could be lost, and the diesel-driven pump might not start because the
lead-acid batteries powering its starter could become disconnected
(the batteries were located on a weakly anchored metal storage rack,
and were not fastened to the rack). Thus, in a seismic event, the
fire main could fail to remain pressurized. At this plant, water was
the agent used in the FPSs for the cable spreading room, the emergency
diesel generator rooms, and many other areas (a seismic event
potentially increases the likelihood of a fire in those and other
critical areas of the plant).
b. The supply reservoir for one CO2 FPS was a non-seismically mounted
tank, and the batteries that supplied power to the tank outlet valve
were weakly anchored to a shelf that had no end restraints. The tank
outlet piping could be damaged and/or valve power could be unavailable
during a seismically induced fire. In this plant, CO2 was the FPS
agent for the cable spreading room, the emergency diesel generator
rooms, and other plant areas.
c. The supply bottles for one Halon FPS were attached to a non-
seismically qualified wall by a single metal strap, providing a high
likelihood that the bottle outlet piping could be damaged and the
Halon could fail to be distributed if demanded by a fire during a
seismic event. In this plant, the Halon was the suppressant agent for
the cable spreading room.
5. Switchgear Fires
Seismic/fire interaction is a contributor to risk in the emergency
electrical distribution rooms due to the presence of a fire source (the
switchgear itself). In some switchgear rooms, many critical cables are
routed along the tops of the switchgear cabinets so that large numbers of
these cables are vulnerable to a fire in any cabinet subdivision. To
reduce the potential risk associated with these areas, some licensees have
implemented the following options:
a. Reduction of fire probability by securing the cabinets with seismic
anchors to prevent tipping or sliding.
.
Attachment 1
IN 94-12
February 9, 1994
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b. Distancing the safety-related cables from the fire source or
separating safety-related equipment cables by distance or physical
barriers.
c. Routing some cables out of the switchgear cabinets through locations
other than the top of the switchgear to reduce the likelihood that a
fire in a single cubicle could damage a large number of safety-related
cables.
6. Electro-Mechanical Components in Cable Spreading Rooms
Many cable spreading rooms contain electrical cabinets, increasing the
risk due to seismic/fire interaction in these rooms. When such cabinets
are present, fire probability can be reduced by securing the cabinets with
seismic anchors to prevent tipping or sliding.
.
Attachment 2
IN 94-12
February 9 1994
Page 1 of 1
REFERENCES
1. J. A. Lambright et al., Risk Evaluation for a Westinghouse Pressurized
Water Reactor, Effects of Fire Protection System Actuation on Safety-
Related Equipment (Evaluation of Generic Issue 57), NUREG/CR-5789,
SAND91-1534, December 1992.
2. J. A. Lambright et al., Risk Evaluation for a Babcock and Wilcox
Pressurized Water Reactor, Effects of Fire Protection System Actuation on
Safety-Related Equipment (Evaluation of Generic Issue 57), NUREG/CR-5790,
SAND91-1535, December 1992.
3. J. A. Lambright et al., Risk Evaluation for a General Electric Boiling
Water Reactor, Effects of Fire Protection System Actuation on
Safety-Related Equipment (Evaluation of Generic Issue 57), NUREG/CR-5791,
SAND91-1536, December 1992.
4. G. Simion et al., Risk Evaluation of a Westinghouse 4-Loop PWR, Effects of
Fire Protection System Actuation on Safety-Related Equipment (Evaluation
of Generic Issue 57), EGG-NTA-9081 Letter Report, Idaho National
Engineering Laboratory, December 1991.
5. J. A. Lambright et al., Evaluation of Generic Issue 57: Effects of Fire
Protection System Actuation on Safety-Related Equipment (Main Report),
NUREG/CR-5580, SAND90-1507, December 1992.
6. Regulatory Analysis for the Resolution of Generic Issue 57: Effects of
Fire Protection System Actuation on Safety-Related Equipment, NUREG-1472,
October 1993.
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