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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 SafetyRelated 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 pressurizedwater 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.

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 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

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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

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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

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 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

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 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.

Page Last Reviewed/Updated Friday, May 22, 2015

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