Information Notice 1994-12, Insights Gained from Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment

Insights Gained from Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment

ML031060630
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	02/09/1994
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
GI-057 IN-94-012, NUDOCS 9402030011
Download: ML031060630 (11)

-

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.

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.- N 94-12 February 9, 1994 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.

Brian K. Grmes, Di tor

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

Reattachment 1 IN 94-12 February 9, 1994 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 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.

• _> K.ttachment 1 IN 94-12 February 9, 1994 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 ina single cubicle could damage a large number of safety-related

cables.

6. Electro-Mechanical Components inCable Spreading Rooms

Many cable spreading rooms contain electrical cabinets, increasing the

risk due to seismic/fire interaction inthese rooms. When such cabinets

are present, fire probability can be reduced by securing the cabinets with

seismic anchors to prevent tipping or sliding.

-> <ttachment 2 IN 94-12 February 9 1994 REFERENCES

1. J. A. Lambright et al., Risk Evaluation for a Westinghouse Pressurized

Water Reactor. Effects of-Fire Protection System Actuation on Safety- Related Eauipment (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 Eouipment (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 Eauipment (Evaluation of Generic Issue 57), NUREG/CR-5791, SAND91-1536, December 1992.

4. G. Simion et al., Risk Evaluation of a Westinghouse 4-Lop 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 Eauipment (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 Eguipment, NUREG-1472, October 1993.

Att.ainent 3 IN 94-12 February 9, 1994 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

94-11 Turbine Overspeed and 02/08/94 All holders of OLs or CPs

Reactor Cooldown during for nuclear power reactors.

Shutdown Evolution

94-10 Failure of Motor-Operated 02/04/94 All holders of OLs or CPs

Valve Electric Power for nuclear power reactors.

Train due to Sheared or

Dislodged Motor Pinion

Gear Key

94-09 Release of Patients with 02/03/94 All U.S. Nuclear Regulatory

Residual Radioactivity Commission medical

from Medical Treatment and licensees.

Control of Areas due to

Presence of Patients Con- taining Radioactivity

Following Implementation

of Revised 10 CFR Part 20

94-08 Potential for Surveil- 01/01/94 All holders of OLs or CPs

lance Testing to Fail for nuclear power reactors.

to Detect an Inoperable

Main Steam Isolation Valve

93-26, Grease Solidification 01/31/94 All holders of OLs or CPs

Supp. 1 Causes Molded-Case for nuclear power reactors.

Circuit Breaker Failure

to Close

94-07 Solubility Criteria for 01/28/94 All byproduct material and

Liquid Effluent Releases fuel cycle licensees with

to Sanitary Sewerage Under the exception of licensees

the Revised 10 CFR Part 20 authorized solely for

sealed sources.

Potential Failure of 01/28/94 All holders of OLs or CPs

Long-Term Emergency for boiling water reactors.

Nitrogen Supply for the

Automatic Depressurization

System Valves

OL = Operating License

CP - Construction Permit

<_j IN 94-12 February 9 1994 FPS actuations. However, in view of the observed large differences in plant- specific characteristics associated with the effects of FPS actuation, plant- specific analyses would be required to identify such 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.

Thus, it is believed that licensee awareness of the insights presented in this

information notice will be beneficial to licensees when performing their IPEEE

program.

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

Inoperabil1ity 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. d.

Brian K.Grimes

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: H. W. Woods, RES P. M. Madden, NRR

(301) 492-3908 (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

• _SEE PREVIOUS CONCURRENCES

OFFICE RPB:ADM* OGCB:DORS* SAIB:DSIR* C/SAIB:DSIR* D/DSIR*

lANE TechEd AKugler HWoods CAder WMinners

DATE 12/30/93 01/04/94 01/06/94 01/10/94 01/10/9 OFFICE C/SPLB:DSSA* D/DSSA* C/OGCB:DORS* D

lAME CMcCracken ACThadani GHMarcus Bl;n_(

DATE ____ 01/21/94 01/24/94 01/26/94 02/,5E-/94 DOCUMENT NAME: 94-12. IN

KJ

VIN 94-XX

February XX, 1994 FPS actuations. However, in view of the observed large differences in plant- specific characteristics associated with the effects of FPS actuation, plant- specific analyses would be required to identify such 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.

Thus, it is believed that licensee awareness of the insights presented in this

information notice will be beneficial to licensees when performing their IPEEE

program.

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, wFeedwater 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.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: H. W. Woods, RES P. M. Madden, NRR

(301) 492-3908 (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

• SEE PREVIOUS CONCURRENCES

OFFICE RPB:ADM* OGCB:DORS* SAIB:DSIR* C/SAIB:DSIR* D/DSIR*

NAME TechEd AKugler HWoods CAder UMinners

DATE 12/30/93 01/04/94 01/06/94 01/10/94 01/10/94 OFFICE C/SPLB:DSSA* D/DSSA* C/OGCB:DORS D/DORS

lAE _CMcCracken ACThadani GHMarcus O ff BKGrimes

DATE 01/21/94 01/24/94 /.24/94 / /94

___ - l

DOCUMENT NAME: FIREPROT.IN IMAPM/

-) _IN 94-XX

February XX, 1994 FPS actuations. However, in view of the observed large differences in plant- specific characteristics associated with the effects of FPS actuation, plant- specific analyses would be required to identify such 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.

Thus, it is believed that licensee awareness of the insights presented in this

information notice will be beneficial to licensees when performing their IPEEE

program.

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.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: H. W. Woods, RES P. M. Madden, NRR

(301) 492-3908 (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

• SEE PREVIOUS CONCURRENCES

OFFICE RPB:ADM* OGCB:DORS SAIB:DSIR:RES C SAIB IR D I

KANE TechEd AKugler 4 I]HWoods CAde C M/ s

DATE 12/30/93 01/64/94 01/06/94 O1/W0/94 1/XR/94 OFFICE C/o_________ D/DSSA C/OGCB:DORS D/DORS

l NAaE 'n

iCr ACThadan1 GHMarcus BKGrimes

lDATE _/ /f/4 // 94 / /94 / /94 DOCMIENT NAME: FIREPROT.IN

jV IN 94-XX

February XX, 1994 The risk reduction estimates, cost/benefit analyses, and other insights gained

during resolution of GSI-57 have shown that consideration of the insights

contained in this information notice (details are provided in Reference 6) can

reduce risk due to FPS actuations. However, in view of the observed large

differences in plant-specific characteristics associated with the effects of

FPS actuation, plant specific analyses would be required to identify such

improvements. Plant specific analyses of the type needed for this purpose are

underway as part of the Individual Plant Examination of External Events

(IPEEE) program, recommended by Generic Letter 88-20, Supplement 4, issued

June 28, 1991. Thus, it is believed that licensee awareness of the insights

presented in this Information Notice will be beneficial to licensees when

performing their IPEEE program.

Related Generic Communications

Information Notice 83-41, Actuation of Fire Suppression System Causing

Inoperability of Safety-Related Equipment"

Information Notice 85-85, OSystems Interaction Event Resulting in Reactor

System Safety Relief Valve Opening Following a Fire-Protection Deluge System

Malfunction"

Information Notice 86-106, Suppplement 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.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Roy Woods, NRR

(301) 492-3908 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

OFFICE RPB:ADM OGCB:DORS SAIB:DSIR:RES C/SAIB:DSIR D/DSIR

NAME TechEd . AKugler RWoods CAder WMinners

DATE 12l/30/9 5 01/ /94 01/ /94 01/ /94 01/ /94 OFFICE EELB:DE:NRR C/EELB:DE D/DE C/OGCB:DORS D/DORS

MAME CBerl nger MWHodges GHMarcus BKGrimes

DATE / /94 / /94 / /94 / /94 / /94 DOCUMENT NAME: FIREPROT.IN