NUREG-1472

Regulatory Analysis for the

Resolution of Generic Issue 57:

Effects of Fire Protection

System Actuation on

Safety-Related Equipment

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

H. W Woods
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NUREG-1472

Regulatory Analysis for the

Resolution of Generic Issue 57:

Effects of Fire Protection

System Actuation on

Safety-Related Equipment

Manuscript Completed: October 1993

Date Published: October 1993

H. W. Woods

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

ABSTRACT

Actuation of Fire Protection Systems (FPS) in Nuclear

Power Plants have resulted in adverse interactions with

equipment important to safety. Precursor operational

experience has shown that 37% of all FPS actuations

damaged some equipment, and 20% of all FPS actuations

have resulted in a plant transient and reactor trip. On an

average, 0.17 FPS actuations per reactor year have been

experienced in nuclear power plants in this country. This

report presents the regulatory analysis for GI-57, "Effects

of Fire Protection System Actuation on Safety-Related

Equipment". The risk reduction estimates, cost/benefit

analyses, and other insights gained during this effort have

shown that implementation of the recommendations contained in this report can significantly reduce risk, and that

these improvements can be warranted in accordance with

the backfit rule, 10 CFR 50.109(a)(3). However, plant

specific analyses are required in order to identify such

improvements. Generic analyses can not serve to identify

improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for

this purpose are underway as part of the Individual Plant

Examination of External Events (IPEEE) program.

iii NUREG-1472

CONTENTS

Page

A bstract ........................................................................................ iii

List of Thbles .................................................................................... vii

Executive Sum m ary .............................................................................. ix

1 Statem ent Of The Problem .................................................................... 1

2 O bjective .................................................................................... 2

3 Alternative Resolutions ....................................................................... 2

3.1 Alternative 1 - No Action ................................................................. 2

3.2 Alternative 2 - Upgrade Relay-Based FPS Actuation Controllers with

Seismically Qualified Printed Circuit Boards ................................................ 2

3.3 Alternative 3 - Upgrade Fire Protection Systems from Smoke Detector

Actuated to Heat Detector Actuated ....................................................... 2

3.4 Alternative 4 - Reroute Safety-Related Cables .............................................. 2

3.5 Alternative 5 - Seismically Qualify CO2 Tlnks, Outlet Piping, and Battery Rack ................. 3

3.6 Alternative 6 - Seismically Qualify FPS Battery Racks ........................................ 3

3.7 Alternative 7 - Upgrade the FPS Water Quality ............. 3

3.8 Alternative 8 - Replace Deluge with Preaction Sprinkler FPS in Selected Plant Locations ........ 3

3.9 Alternative 9 - Upgrade Electrical Cabinet Design to Prevent Water Intrusion .................. 3

3.10 Alternative 10 - Seismically Anchor Safety-Related Cabinets Susceptible

to Tipping/Sliding Failures ................................................................ 3

4 Technical Findings ............................................................................ 3

4.1 Core Damage Frequency Analysis .......................................................... 4

4.1.1 Generic FPS Root-Cause Actuation Scenarios ......................................... 4

4.1.2 G eneric Plant Analysis ............................................................. 5

4.1.3 Generic Core Damage Frequency Estimates .......................................... 7

4.2 Dose Consequences Analysis .............................................................. 10

4.3 Cost A nalysis ............................................................................ 11

4.3.1 Labor and Equipment/Materials Costs ............................................... 11

4.3.2 Engineering and Quality Assurance/Control Costs ..................................... 12

4.3.3 Radiation Exposure ................................................................ 12

4.3.4 Health Physics Support Costs ........................................................ 12

4.3.5 Anti-Contamination Clothing Costs .................................................. 13

4.3.6 Radioactive Waste Disposal Cost ..................................................... 13

4.3.7 Other Licensee Costs .............................................................. 13

4.3.8 N RC Costs ........................................................................ 13

4.3.9 Onsite Averted Costs ............................................................... 13

4.4 Cost Estimate Uncertainties ............................................................... 14

4.5 Backfit Alternatives Cost Estimates ........................................................ 14

4.5.1 Alternative 1 - No Action ........................................................... 14

4.5.2 Alternative 2 - Upgrade a FPS Actuation Controller with Seismically Qualified

Printed Circuit Boards .................................................. 14

4.5.3 Alternative 3 - Replace Smoke Detector Actuated FPS with Heat Detector

A ctuated System ................................................................... 14

4.5.4 Alternative 4 - Reroute Safety-Related Cables ........................................ 14

4.5.5 Alternative 5 - Seismically Qualify the C02 Thnk, Outlet Piping, and Battery Rack ......... 15

v NUREG-1472
CONTENTS (Continued)

Page

4.5.6 Alternative 6 - Seismically Qualify A FPS Battery Rack ................................ 15

4.5.7 Alternative 7 - Upgrade the FPS Water Quality ................................. 15

4.5.8 Alternative 8 - Replace Deluge with Preaction Sprinkler FPS 15

4.5.9 Alternative 9 - Replace Electrical Cabinet with a Cabinet Designed to Prevent

Water Intrusion ....................................................... 16

4.5.10 Alternative 10 - Seismically Anchor Safety-Related Cabinets Susceptible to

Tipping/Sliding Failure ...................................... ....................... 16

5 Value/Impact Analysis ......................................................... 17

5.1 Backfit Analysis ...................................................................... ;.... 17

5.1.1 Alternative 1 - No Action ............................................................ 17

5.1.2 Alternative 2 - Upgrade an FPS Actuation Controller with Seismically Qualified

Printed Circuit Boards ........................... IT

5.1.3 Alternative 3 - Replace Smoke Detector Actuated FPS with a Heat Detector

Actuated System ................... 17

5.1.4 Alternative 4 - Reroute Safety-Related Cables ................................. 17

5.1.5 Alternative 5 - Seismically Qualify the CO 2 Tank, Outlet Piping, and Battery Rack .......... 17

5.1.6 Alternative 6 - Seismically Qualify A FPS Battery Rack ......................... ....... 18

5.1.7 Alternative 7 - Upgrade the FPS Water Quality ................................. 18

5.1.8 Alternative 8 - Replace Deluge with Preaction Sprinkler FPS .......... ............. 20

5.1.9 Alternative 9 - Replace Electrical Cabinet with a Cabinet Designed to Prevent

W ater Intrusion .................................................................... 20

5.1.10 Alternative 10 - Seismically Anchor Safety-Related Cabinets Susceptible to

Tipping/Sliding Failure .................................................. 22

5.2 Frontfit Analysis .......................................................................... 22

5.2.1 Alternative 1 - No Action ............................................... 23

5.2.2 Alternative 2 - Upgrade a FPS Actuation Controller with Seismically Qualified

Printed Circuit Boards .................................................. 23

5.2.3 Alternative 3 - Replace Smoke Detector Actuated FPS with a Heat Detector

Actuated FPS ..................................................................... 23

5.2.4 Alternative 4 - Reroute safety-related Cables ......................................... 24

5.2.5 Alternative 5 - Seismically Qualify the CO2 1Tank, Outlet Piping, and Battery Rack ........ 24

5.2.6 Alternative 6 - Seismically Qualify a FPS Battery Rack ................................. 24

5.2.7 Alternative 7 - Upgrade the FPS Water Quality ....................................... 24

5.2.8 Alternative 8 - Replace Deluge with Preaction Sprinkler FPS ........................... 24

5.2.9 Alternative 9 - Replace Electrical Cabinet with a Cabinet Designed to Prevent

Water Intrusion ..................................... ................. 24

5.2.10 Alternative 10 - Seismically Anchor Safety-Related Cabinets Susceptible to

Tipping/Sliding Failure ............................................................. 24

6 D ecision Rationale ........................................................................... 25

7 Im plem entation .............................................................................. 27

8 References .................................................................................. 27

NUREG-1472 vi
LIST OF TABLES

Page

4.1 Fire Protection System Summary. ................................................. 6

4.2 Fire Protection Cases for Analysis .... ,....................... ............................... . 6

4.3 Fire Occurrence. Frequencies .. ......................... 6

4.4 FPS Actuation Frequency per System-Year ................................................... 7

4.5 Core Damage Frequency -,"Most Vulnerable" Generic Plant .... ................................. 8

4;6 Core Damage Frequency - "Typical" Generic Plant .................................... 9

4.7 Core Damage Frequency - "Least Vulnerable" Generic Plant ............................... 9

4.8 Plant-Specific Base Case Results in Terms of Core Damage Frequency (Per Reactor Year) ......... 10

4.9 Risk in person-Sv/RY (person-rem/RY) "Typical" PWR Plant ................................... 11

4.10 Risk in person-Sv/RY (person-rem/RY) "Typical" BWR Plant ................................... 12

5.1, Alternative 2 $K/person-Sv ($K/person-rem) averted................................... 18

5.2 Cost/Benefit Ratio Alternative 3 $K/person-SV ($K/person-rem) averted ..................... 19

5.3 Cost/Benefit Ratio Alternative 4 $K/person-Sv ($K/person-rem) averted ......................... 19

5.4. Cost/Benefit Ratio Alternative 5 $K/person-Sv ($K/person-rem) averted ........................ 20

5.5 Cost/Benefit Ratio Alternative 6 $K/person-Sv ($K/person-rem) averted ........................ 21

5.6: Cost/Benefit Ratio Alternative 8 $K/person-Sv ($K/person-rem) averted ..................... 21

5.7 Cost/Benefit Ratio Alternative 9 $K/person-Sv,($K/person-rem) averted ....... .............. 22

5.8 Cost/Benefit Ratio Alternative 10 $K/person-Sv ($K/person-rem) averted.................... 23

vii . NUREG-1472

EXECUTIVE SUMMARY

This report provides supporting information, including a

value/impact analysis for the Nuclear Regulatory Commission's (NRC's) resolution of Generic Issue 57, "Effects of Fire Protection System Actuation on SafetyRelated Equipment". This issue addresses the concerns

regarding the actuation of fire protection systems (FPS)

using water, carbon dioxide and halon, and the effects of

these fire suppressants on plant equipment, the damage

or upset of which might initiate and sustain risk significant

event sequences.

FPS actuations that result in adverse interactions with

plant systems needed to achieve safe plant shutdown or to

mitigate a postulated accident reduce the availability of

such systems. This concern is accentuated when common

cause initiators and common mode failures of safetyrelated equipment are considered. Examples of common

cause initiators include earthquakes, smoke and heat intrusion into multiple fire zones, and fire suppressant intrusion into multiple fire zones affecting several safetyrelated systems. Examples of common mode failures of

safety-related systems and/or auxiliary systems supporting

safety-related systems include electrical shorts in instrument cabinets and electrical power distribution centers,

CO2 ingress into the fresh air intake of emergency dieselgenerator sets, and cold CO2 induced thermal stresses

and cracking of station battery casings, with loss of offsite

power during an earthquake. It should be noted that a

number of common cause initiators and common mode

failures are not mutually exclusive and they may be part of

a single event sequence.

Generic Issue 57 was identified in 1982 (Refs. 1,2) as a

result of a number of precursor events showing that safety-related equipment subjected to fire protection system

(FPS) water spray could be rendered inoperable. These

precursor events also indicated numerous spurious FPS

actuations initiated by operator testing errors or by maintenance activities, e.g., welding, and steam or high humidity in the vicinity of FPS detectors or control circuitry and

components. An NRC memorandum (Ref.2), issued on

January 28, 1982, provided additional examples of FPS

actuation interactions and suggested that all types of FPS

be included in a review of their safety significance and

possible corrective steps.

On June 22, 1983, IE Information Notice 83-41 (Ref.3)

was issued to alert licensees and to provide examples of

recent experiences in which actuation of fire protection

systems caused damage or inoperability of safety-related

systems. The effects of such events range from reactor

trips to fires in high voltage electrical equipment, and

water contamination of diesel fuel. The IE Information

Notice indicated that the plant Fire Hazards Analysis

under Appendix R to 10 CFR 50 requires, not only consideration of the consequences of a postulated fire, but also

consideration of the effects of fire-fighting activities. The

IE Information Notice also stated that a properly conducted Fire Hazards Analysis in conjunction with a physical walkdown of plant areas would have identified instances where minor modifications such as shielding

equipment and sealing conduit ends would have reduced

equipment water damage from inadvertent FPS operation.

Four plants representative of the various designs of currently operating nuclear power plants were evaluated as

part of this issue (Refs. 4-8). Furthermore, a generic

evaluation of this issue (Ref. 9) was performed taking into

account the insights from the technical findings of these

four evaluations, as well as design and plant layout information of a large number of operating plants collected for

this purpose. An extensive review of operational experience involving actuations of fire protection systems was

performed prior to the analytical assessments of risk associated with this issue. The review of the operational experience showed the following:

* 0.15 inadvertent FPS actuations/RY

* 0.02 advertent FPS actuations/RY

S

S

37% of all actuations damaged some equipment

20% of all actuations resulted in a plant transient

and reactor trip

" Core Damage Frequency (CDF) contributions from

GI-57 root causes for the four individual plants evaluated were estimated to be in the range of

7.3E-06/RY to 5.6E-05/RY.

* Dominant risk contributors are associated with seismic/FPS and seismic/fire interactions resulting in sequences involving station blackout and small

LOCAs

Both of these categories of dominant sequences are currently being addressed by the Individual Plant Examination of External Events (IPEEE) program. Generic Letter 88-20, Supplement 4, June 27, 1991 (which initiated

the IPEEE) states:

"The walkdown procedures should be specifically

tailored to assess the remaining issues identified in

the Fire Risk Scoping Study: (1) seismic/fire interactions, (2) effects of fire suppressants on safety equipment, .... " (paragraph 4.2).

The staff notes that any complete assessment of seismic/

fire interactions would necessarily include seismic effects

on manual firefighting, since automatic fire suppressant

systems are not seismically qualified, thus increasing the

ix NUREG-1472
potential need for effective manual firefighting during

seismic events.

NUREG-1407 ("Procedural and Submittal Guidance for

the IPEEE for Severe Accident Vulnerabilities", June,

1991) reiterates the above (in the first paragraph of section 4) and also adds:

"The use of an existing fire PRA for the internal fires

IPEEE is acceptable provided the PRA reflects the

current as-built and as-operated status of the plant

and the licensee addresses the deficiencies of past

PRAs that are identified in the Fire Risk Scoping

Study (NUREG/CR-5088). Deficiencies may include the use of low conditional failure probabilities

for dampers and penetrations, no consideration of

damage from the use of fire suppressants, inappropriate estimates of the effectiveness of manual fire

fighting, and no consideration of seismic/fire interactions." (paragraph 4.2).

Also in NUREG-1407, Appendix D, section 6 ("Internal

Fires") the staff response to question 6.2 states:

'The procedurally directed walk-downs associated

with internal fires vulnerability evaluation can be

planned as part of the seismic walk-downs that would

specifically look for the seismic-induced fire vulnerability issues. The idea is to first identify those areas

that could be vulnerable so that they can be brought

into focus during the walkdown.

"For example, if a plant didn't have its diesel fuel

tank strapped down properly one could postulate a

large fuel source for fire as a result of a seismic

event. Other similar seismic/fire interactions were

summarized in Section 7 of NUREG/CR-5088."

In addition, for performance of the IPEEE, many licensees will also be using the Fire-Induced Vulnerability Evaluation (FIVE) methodology developed by the Electric

Power Research Institute (EPRI) as described in "Fire-Induced Vulnerability Evaluation (FIVE)", EPRI TR100370, April, 1992. Following a description of the three

basic Phases of FIVE, that document states:

In addition, there is a discussion of several potentially risk significant items that were identified in the

NUREG/CR-5088, "Fire Risk Scoping Study", that

should also be considered in performing FIVE. (Section 7.0) [emphasis added].

Section 7.0 of NUREG/CR-5088 is a description of the

Sandia Fire Risk Scoping Study Evaluation which includes

discussion of the dominant risk contributors to GSI-57-

related events, including Seismic/Fire interactions (seismically induced fires, seismic actuation of fire suppression

systems, and seismic degradation of fire suppressant systems), and Manual Fire Fighting Effectiveness. The following seismic/fire interactions are given as exam.pies 0'of

situations to be considered by the IPEEE: :unarnlchored`''

CO2 or Halon tanks, possible relay chatter in fiie protection system actuation systems, fire alarm systems having

only a smoiketactuated alarm without a heat'or fla•ie.

detector, fire pump mounts without vibration ampiitude`ýstops, cast iron fire mains, inadequate anchoring of electrical cabinets and inadequate slack in the wires leading to

such cabinets (to avoid sparks from tight wires), unanchored high-pressure gas bottles, potential interactions

between sprinkler system heads and adjacent pipes, and

presence of mercury switches in fire suppression and detection systems (such switches should be replaced with

alternate, seismically insensitive switches).

For the "generic" evaluation, after subtraction of CDF

contributions from GSI-57-related events involving:

(a) Seismic/Fire: - seismic induced fire plus seismic induced suppressant diversion. The unsuppressed fire and/or the diverted suppressant

incapacitate safety related equipment needed

to mitigate effects of the seismic event; and

(b) Seismic/FPS: - seismic induced actuation of the

FPS. Released suppressant damages safetyrelated equipment needed to mitigate effects

of the seismic event,

which are being emphasized by the IPEEE, the mean

CDF of the remaining contributors is less than

1.OE-05/RY which does not justify a generic backlfit.

The risk reduction estimates, cost/benefit analyses, and

other insights gained during this effort have shown that

implementation of the recommendations contained in

this report can significantly reduce risk, and that these

improvements can be warranted in accordance with the

backfit rule, 10 CFR 50.109(a)(3). However, plant specific

analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for

this purpose are underway as part of the Individual Plant

Examination of External Events (IPEEE) program.

Hence, we concluded that GSI-57 should be resolved with

no new regulatory requirements generically imposed for

existing plants.

For new plants, the enhanced safety requirements given

in SECY-90-016, "Evolutionary Light Water Reactor

(LWR) Certification Issues and their Relationship to Current Regulatory Requirements", January 12, 1990, allow

the same GSI-57 resolution to be applied (i.e., with no

new regulatory requirements imposed). This is justified

since the designers of new plants will perform

NUREG-1472 X
plant-specific PRAs, including external events, that are

equivalent to the IPEs and EPEEEs which will resolve this

issue for existing plants.

However, the reports dealing with the evaluation of this

issue, including this regulatory analysis, will be published,

or otherwise be made publicly available, so that the insights gained from this effort may be used by licensees and

applicants, and the nuclear industry in general, to take

voluntary steps for plant-specific cases, including frontfitting considerations for advanced or evolutionary reactor

designs.

xi xi NUREG-1472
.4
1 STATEMENT OF THE PROBLEM

Generic Issue 57, "Effects of Fire Protection System Actuation on Safety-Related Equipment", was identified in

1982 (Refs. 1,2) as a result of a number of precursor events

showing that safety-related equipment subjected to fire

protection system (FPS) water spray could be rendered

inoperable. These precursor events also indicated numerous spurious FPS actuations initiated by operator testing

errors or by maintenance activities, e.g., welding, and

steam or high humidity in the vicinity of FPS detectors or

control circuitry and components. An NRC memorandum

(Ref.2), issued on January 28, 1982, provided additional

examples of FPS actuation interactions and suggested

that all types of FPS, e.g., water, halon, carbon dioxide

and other chemicals be included in a review of their safety

significance and possible corrective steps.

A review of FPS related regulations and guidelines regarding interactions between FPS features and plant safety systems, as well as a review of operating experience

.involving such interactions, led to the conclusion that, if

existing requirements are properly implemented, such

interactions should be minimized. GDC-3 states that inadvertent FPS operation or failure should not impair the

function of safety systems. However, licensee event reports (LERs) indicated that regulatory requirements, not

necessarily limited to GDC-3, had not been properly implemented at some plants. For example, deficiencies in

sealing of electrical cable conduits resulted in the transport of water in instrument and electrical cabinets resulting in equipment damage or upset including FPS actuations.

On June 22, 1983, IE Information Notice 83-41 (Ref.3)

was issued to alert licensees and to provide examples of

recent experiences in which actuation of fire protection

systems caused damage or inoperability of safety-related

systems. The effects of such events range from reactor

trips to fires in high voltage electrical equipment, and

water contamination of diesel fuel. The IE Information

Notice indicated that the plant Fire Hazards Analysis

under Appendix R to 10 CFR 50 requires, not only consideration of the consequences of a postulated fire, but also

consideration of the effects of fire-fighting activities. The

IE Information Notice also stated that a properly conducted Fire Hazards Analysis in conjunction with a physical walkdown of plant areas would have identified instances where minor modifications such as shielding

equipment and sealing conduit ends would have reduced

equipment water damage from inadvertent FPS operation. The IE Information Notice indicated that none of

the reported events resulted in a serious impact on the

functional capability of a plant to protect public health

and safety. However, examples were given where it would

not be difficult to extrapolate actual occurrences into a

sequence of events that could lead to more serious consequences.

FPS actuations that result in adverse interactions with

plant systems needed to achieve safe plant shutdown or to

mitigate a postulated accident reduce the availability of

such systems. This concern is accentuated when common

cause initiators and common mode failures of safetyrelated equipment are considered. Examples of common

cause initiators include earthquakes, smoke and heat intrusion into multiple fire zones, and fire suppressant intrusion into multiple fire zones affecting several safetyrelated systems. Examples of common mode failures of

safety-related systems and/or auxiliary systems supporting safety-related systems include electrical shorts in instrument cabinets and electrical power distribution centers, CO2 ingress into the fresh air intake of emergency

diesel-generator sets, and cold CO2 induced thermal

stresses and cracking of station battery casings, with loss

of offsite power during an earthquake. It should be noted

that a number of common cause initiators and common

mode failures are not mutually exclusive and they may be

part of a single event sequence.

Four plants representative of the various designs of currently operating nuclear power plants were evaluated as

part of this issue (Refs. 4-8). Furthermore, a generic evaluation of this issue (Ref. 9) was performed taking into

account the insights from the technical findings of these

four evaluations, as well as design and plant layout information of a large number of operating plants collected for

this purpose. An extensive review of operational experience involving actuations of fire protection systems was

performed prior to the analytical assessments of risk associated with this issue. The review of the operational experience showed the following:

* 0.15 inadvertent FPS actuations/RY

* 0.02 advertent FPS actuations/RY

* 37% of all actuations damaged some equipment

* 20% of all actuations resulted in a plant transient

and reactor trip

* Core Damage Frequency (CDF) contributions from

GI-57 root causes for the four individual plants evaluated were estimated to be in the range of

7.3E-06/RY to 5.6E-05/RY.

0 Dominant risk contributors are associated with seismic/FPS and seismic/fire interactions resulting in sequences involving station blackout and small

LOCAs

I NUREG-1472
2 OBJECTIVE

The objective of the Generic Issue 57 resolution program

is to evaluate the present design of fire protection systems

in operating nuclear power plants and the effects of their

actuation on plant safety, and to examine the cost effectiveness of alternative measures for reducing the overall

vulnerability of safety systems to FPS actuations. It is also

the objective of this issue to assess these alternative corrective measures for their applicability and efficacy to

evolutionary or new designs of nuclear power plants.

Probabilistic methods were used to assess the CDF, the

potential reduction in risk' of alternative corrective measures and their cost-effectiveness. The overall objective of

the resolution of GI-57 is that contribution from FPS

actuations should be a small percentage of the total CDF

from all causes.

For USI A-45, the staff recommended that the frequency

of events related to decay heat removal failure leading to

core damage should be reduced to a level (about

1.OE-05/RY) that the probability ofi S'ucch anr accident in

the next 30 years would be about 0.03 based on a population of about 110 plants. A CDF objective of 1.0E-05/RY.

was also noted in USI A-44, "Station Blackout." These

objectives are also. consistent with the recently issuedd

guidance to the staff (Ref. 10) setting a safety goal of less

than 1.0E-04/RY for CDF from all contributors, with a

subsidiary safety goal of no more than 1.OE-05/RY fora

single contributor. The interim guidance provided in Reference 10 for purposes of satisfying the substantial additional protection criterion of the backfit rule setsforth-the,

guideline that initiatives involving actions to reduce CDF

should result in at least 1.0E-05/RY reduction'. In assessing the risk reduction significance of potential backfits the,

conditional containment failure probability (CCFP) musvt

also be evaluated.against the guidance provided in Refer-,

ence 10. Section 6 of this regulatory analysis takes into

account all these considerations, including attendant uncertainty and sensitivity analyses in developing the decision rationale for resolving this issue.

3 ALTERNATIVE RESOLUTIONS

There were several alternatives analyzed for the resolution of Generic Issue 57. These alternatives are described

below:

3.1 Alternative 1 - No Action

With this alternative, there would be no regulatory requirements. Consistent with existing regulations, this alternative does not preclude a licensee, or an applicant for

an operating license, from proposing to the NRC staff

design changes intended to enhance the reliability or operability of the protection systems and their components

on a plant-specific basis.

3.2 Alternative 2 - Upgrade RelayBased FPS Actuation Controllers

with Seismically Qualified Printed

Circuit Boards

During our evaluation of GI-57 it was determined that

many types of relays are used for the actuation of fire

protection systems. In some plants, mercury wetted relays

for fire protection system actuation, alarm annunciation

and isolation of equipment room cooling are used. In the

event of an earthquake it is highly likely that some of these

relay types, and certainly the mercury-wetted ones would

actuate. Under this alternative, relay-based controllers

would be replaced with printed circuit boards to reduce

the potential for inadvertent actuation of fire protection

systems during earthquakes.

3.3 Alternative 3 - Upgrade Fire

Protection Systems from Smoke

Detector Actuated to Heat

Detector Actuated

Smoke detector actuated systems are subject, to more

frequent inadvertent actuations. Modifying such systems'

to heat detector actuated systems would result in reducing

the frequency of inadvertent FPS actuations. It might be,

prudent to retain the existing smoke detectors for indication or alarming only, or as a part of a coincident logic

actuation system based on heat and smoke detectors.

3.4 Alternative 4 - Reroute

Safety-Related Cables

Our evaluation of this issue, including some plant walk-,

downs, found that there are certain "pinch points" in

plants where cables for some redundant safety-related

systems are routed close together. Given a fire or FPS

actuation which would damage these cables, the respective safety-related systems would be vulnerable to con-,

mon failure. Hence, the purpose of this alternative would

be to reroute one of the sets of cables to remove this.

common mode failure vulnerability.

NUREG-1472 2
3.5 Alternative 5 - Seismically Qualify

CO2 Thnks, Outlet Piping, and

Battery Rack

A seismic walkdown and subsequent fragility analysis for a

typical CO2 system found a high likelihood of suppressant

agent diversion given an earthquake. Failure of the tank

or its outlet piping or the FPS battery dominated the

overall probability of failure of the CO2 system. This potential plant modification would seismically qualify the

CO2 tank, battery, and its immediate outlet piping.

3.6 Alternative 6 - Seismically Qualify

FPS Battery Racks

It was found during a plant specific walkdown, that a water

based FPS for an entire plant had two electric driven fire

pumps and one diesel driven fire pump. Given a LOSP,

both electric driven pumps would fail due to being powered from non-vital busses. A failure of the starting battery for the diesel pump would lead to a loss of fire main

pressure.

3.7 Alternative 7 - Upgrade the FPS

Water Quality

Water is the most frequently used FPS agent at nuclear

power plants. The most common delivery systems are

pre-action sprinklers, deluge wet-pipe sprinklers, drypipe sprinklers and standpipe hose systems. It has been

postulated that improvement of water quality will reduce

the potential for damage to safety-related components

from exposure to water suppressant. To address this concern, a modification to upgrade the FPS water quality is

performed. For this modification a water purification system would be required along with the associated piping

and valves and a storage tank. It is assumed for this modification that the existing FPS piping and pumps would be

utilized.

3.8 Alternative 8 - Replace Deluge

with Preaction Sprinkler FPS in

Selected Plant Locations

Of all of the types of water based FPSs utilized in nuclear

power plants the pre-action sprinkler system has been

found to experience the least number of inadvertent actuations (Appendix A). A pre-action sprinkler system requires the opening of a deluge value, either automatically,

with a control signal, or manually, and a rise in the ambient temperature to the melting point of the fusible links

on the sealed sprinkler heads. A deluge FPS only requires

the opening of a valve, either automatically or manually to

discharge water under pressure to the open spray heads.

This modification involves replacing a deluge FPS with a

pre-action sprinkler FPS. This modification would reduce

the frequency of inadvertent actuations and localize the

application of FPS agent.

3.9 Alternative 9 - Upgrade Electrical

Cabinet Design to Prevent Water

Intrusion

One study (Ref. 11) has reported that electrical equipment failure modes related to water weremainly due to

electrical shorting and long term corrosion. Wherever

water intrusion is possible the equipment could be expected to fail through shorting, grounding, tripping of

overcurrent devices, physical damage due to the velocity

of direct hose streams or long term corrosion causing

potential failure of electro-mechanical parts. These failure modes were found to be dependent on the National

Electrical Manufacturers Association (NEMA) rating or

the configuration of the electrical enclosure. The appropriate NEMA rating to preclude water intrusion can potentially eliminate failures in electrical equipment due to

water based FPSs.

3.10 Alternative 10 - Seismically

Anchor Safety-Related Cabinets

Susceptible to Tipping/Sliding

Failures

In a seismic event energized cabinets present potential

sources of fire. Although it is assumed that in a seismic

event offsite power will be lost, thus deenergizing many

electrical cabinets, there will be a number of safetyrelated cabinets energized by alternative power sources

(batteries, diesel generators). These energized cabinets

are susceptible to tipping/sliding failure possibly leading

to a fire.

It is assumed that this modification would eliminate the

potential for seismically induced fires due to the tipping/

sliding failure of an energized electrical cabinet.

4 TECHNICAL FINDINGS

Sandia National Laboratories (SNL) was the primary contractor for performing the technical evaluation of the

effects of fire protection system actuation on safetyrelated equipment. This technical evaluation included a

determination of the frequency of FPS actuations, core

damage frequency from accident sequences resulting

3 NUREG-1472
from FPS actuations, dose consequence analysis, and

cost/benefit analysis. The detailed evaluations of specific

plant designs are presented in References 4-8 and the

generic, summary evaluation report is contained in Reference 9, which also contains an evaluation of the operational experience involving FPS actuations and the SNL

methodology of performing the vital area and seismic

analyses. This discussion of the technical findings is a

condensed presentation drawn from the above references

and Reference 9 in particular.

4.1 Core Damage Frequency Analysis

The core damage vulnerability caused by FPS actuations

with the potential of causing damage to safety-related

equipment was estimated using existing PRA models. A

general transient event tree was used to account for FPS

actuations. For seismically-induced FPS actuations, the

loss of offsite power (LOSP) transient event tree was

used. The success or failure of the required safety systems

was determined by fault tree analyses identifying logically

all possible combinations of component failures leading to

the failure of the safety system in question. These logic

models were combined using Boolean algebra as embodied in the SETS (Ref. 12) computer code, ultimately

leading to the derivation of the "minimal cut sets" and

accident sequences.

4.1.1 Generic FPS Root-Cause Actuation

Scenarios

Based on the review of past experiences and walkdowns of

a number of plants, thirteen generic root-cause scenarios,

as shown in Thble 3.1, were identified. Three root causes

are due to inadvertent FPS actuations caused by a fire in

another zone. Four are due to actuations resulting from

purely random causes. Four are due to seismic causes, and

one is due to the occurrence of a fire outside the plant.

Also included is advertent FPS actuation with the presence of a fire. The various root causes of FPS actuation

are described below and the specific tasks and information

required to evaluate them are briefly discussed:

1. Fire-induced FPS actuation - FPS agent-induced

,darnag. Based on the vital area analysis and plant

specific (for example data submitted in accordance

with 10CFR50 Appendix R), fire zones are identified

where smoke or heat spread could cause inadvertent

actuation in other plant areas which are either physically adjacent or connected through ventilation

paths. Estimates are made of the impact of the FPS

agent on equipment in these plant areas and are

applied to the appropriate cut sets and accident sequences.

2. Fire-induced FPS actuation - recovery prevention.

A plant's fire PRA is reviewed for risk-significant

recovery actions on equipment not damaged by fire

and it is also determined in which plant areas these

actions must occur. Then, other fire zones which are

either physically adjacent or connected through ventilation are examined to determine if either heat or

smoke spread could actuate the FPS and prevent the

recovery action hypothesized. If any such combinations are found, the applicable accident sequences

are requantified.

3. Fire-induced FPS actuation - access prevention. For

each critical fire zone identified in the fire PRA

where manual fire suppression was identified as the

means of mitigating the fire, access to the fire zone is

identified via plant specific data and a plant walkdown. As was the case for smoke or heat spread

actuating a FPS and preventing recovery actions, a

similar analysis is conducted for the delay in manual

fire suppression caused by FPS actuation and the

applicable accident sequences are requantified.

4. FPS actuation - human error. The vital area analysis

and plant data are reviewed to determine which fire

zones have an FPS that can be manually actuated.

Erroneous manual actuation in any singie fire zone

may occur due to false detector signals or human

errors of commission. For those cut sets requiring

failures in more than one fire zone, the most likely

scenario is that of a fire in one of the areas. Frequencies for such events are obtained from the historical

data base for different types of FPSs. Using these

frequencies, the accident sequences are requantified.

5. FPS actuation - steam pipe break. This root cause

quantifies the core damage frequency contribution

from inadvertent FPS actuation caused by a hightemperature steam environment. This actuation can

occur due to moisture intrusion into a FPS controller, activation of smoke/heat detector(s), or melting

of fusible link heads. An estimate of steam release

frequency is made and the applicable accident sequences are requantified.

6. FPS actuation - hardware failures of FPS. In this

scenario, inadvertent actuation of the FPS is caused

by hardware failures of the FPS itself, such as a pipe

break in a wet pipe system, or a failure in an FPS

control circuit. Frequencies for such events are obtained from the historical data base for different

types of FPSs. Using these frequencies, the accident

sequences are requantified.

7. Seismic FPS actuation - dust. Those fire zones

where automatic FPS are actuated solely by smoke/

particulate detectors are identified. Then dust is assumed to cause FPS actuation in the fire zone given a

seismic occurrence. The additional FPS damaged

NUREG-1472 4
equipment failures are added to the seismic sequences and these sequences are then requantified.

8. Seismic FPS actuation - relay chatter. The potential

for seismically induced relay chatter is quantified

based on a detailed evaluation of each FPS actuation

circuit within a given plant. The additional FPS damaged equipment failures are added to the seismic

sequences and these sequences are then requantified.

9. Seismic FPS actuation - mechanical failures. The

potential for seismically induced mechanical failure

is quantified based on a detailed evaluation and a

plant walkdown of each FPS. The vital area equations can again be used directly to assess the impact

of such events.

10. External fire-caused FPS actuation. Frequency of

smoke intake from external fires is estimated from a

combination of generic and plant-specific data. Fire

zones potentially affected by smoke spread from outside ventilation are identified.

11. FPS actuation with the presence of a fire. Quantification of this scenario requires either an existing fire

PRA or identification of fire sources in critical fire

zones. Each fire zone with a FPS is identified to

judge the effect of a fire with the simultaneous release of FPS agent on otherwise undamaged vital

equipment. These failures are added to the fire sequences and requantified.

12. Seismic/fire interaction. In this scenario one or more

seismically induced fires are evaluated for the probability of occurrence based on a plant walkdown and

seismic fragility analysis of fire sources within the

zone(s). The probability of diversion of FPS agents

into zones not containing the fire(s) is made. These

failures are added to the seismic sequences and requantified.

13. FPS actuation - unknown causes. In this scenario

inadvertent actuation of the FPS is due to unknown

causes. Frequencies for such events are obtained

from the historical data base for different types of

FPS. Using these frequencies, the accident sequences are requantified.

Tb identify the critical plant fire zones, criteria were developed for each root cause scenario which enable the

analyst to determine which zones are potentially subject

to each root cause of FPS actuation, given the general

vital area analysis accident sequence equations. These

criteria are shown in Tible 3.2 of Reference 9. This step

was performed manually, and required a review of plant

systems, plant layouts, and plant specific data. This review

considered such factors as the following:

" the presence of automatic or manual fixed FPSs,

" physical and electrical separation of redundant

trains,

" susceptibility to seismic events,

* propagation of combustion products (generated either inside or outside the plant) through the ventilation system,

* possible water and steam ingress into vulnerable

equipment,

0

S

single random actuations of FPSs,

multiple actuations of FPSs, and

* type of fire detectors.

4.1.2 Generic Plant Analysis

The safety significance of fire protection system (FPS)

actuation with subsequent safety-related system damage

is highly plant dependent (this is one reason why the

IPEEE program was initiated). However, there are sufficient similarities among plants that some generic conclusions can be drawn. For this generic plant analysis, three

types of fire zones were assessed: the cable spreading

room(s), the emergency electrical switchgear room(s),

and the diesel generator room(s). These zones were selected for several reasons. First, they are all represented

in each of the individual plants studied (Refs. 4-8). Second, in each study, these zones are all contributors to

overall risk. Third, these zones are representative of

equivalent zones in all the U.S. commercial power plants.

More directly, all power plants have spaces that are functional equivalents to cable spreading rooms, diesel generator rooms, and emergency electrical switchgear rooms.

However, it must be emphasized that there may well be, in

specific plants, other fire zones that dominate risk associated with fire suppressant damage to safety-related systems. Individual plant analysis must be done to identify

such zones.

Fire suppressants studied include water, Halon and CO2.

Sensors used in the actuation control systems include:

- Smoke detectors, generally the ionization type.

These sensors are sensitive to dust generated in

a seismic event as well as fire-generated smoke.

- Flame detectors, used primarily in diesel generator rooms and in other areas where oil fires

are likely.

- IR heat detectors.

- Cable tray heat detectors, that have long wire

sensors in individual cable trays.

FPS piping systems are of two types; with pressurized and

non-pressurized headers. In the generic cases presented,

5 NUREG-1472
only the wet pipe water system is of the pressurized type.

Water preaction, water deluge, Halon and CO2 are the

non-pressurized type, relying on sensors and the control

system to detect the need for suppressant and subsequently actuate pressurization valves. Thble 4.1 summarizes the elements of the various types of systems analyzed

in the generic cases.

Different strategies have been employed in U.S. commercial nuclear power plants for critical fire zones as to the

type of suppressant agent and fire protection system actuation scheme utilized. Table 4.2 lists the cases that were

analyzed in this study.

Table 4.1 Fire Protection System Summary

Normally Sensors Nozzle

System Type Pressurized Utilized Type

H 20 Preaction No Yes Fusible Link

H 20 Wetpipe Yes No Fusible Link

H20 Deluge No Yes Open

Halon No Yes Open

CO 2 No Yes Open

Table 4.2 Fire Protection Cases for Analysis

Cable Diesel Emergency

Spreading Generator Switchgear

Suppression Agent System Type Room Rooms Rooms

Water Preaction X X

Water Wetpipe X X X

Water Deluge X X X

Halon X - X

CO 2 X X X

X indicates analysis was performed

The conditions and assumptions used in the generic scenario analysis are discussed in detail in Reference 9. A

data base of fire occurrences was developed in Reference

13 for a number of typical nuclear power plant fire zones.

Those applicable to this generic issue are presented in

Table 4.3.

Fragility data for seismic suppressant diversion are based

on individual plant specific analysis. It must be noted that

while diversion is a concern in the assessment of Root

Cause 12, it acts as an advantage in the analysis of Root

Causes 7/8/9, because if suppressant is diverted away from

a zone of interest, it cannot cause damage to safetyrelated components and cables.

Non-seismic fire protection system actuation frequencies

(per system-year) are based on the LER data in Appendix

A, and are provided in Table 4.4.

Table 4.3 Fire Occurrence Frequencies

Fire Zone Frequency/Reactor-Year

Cable Spreading Room 2.7E-3

Diesel Generator Building 2.3E-2

Emergency Switchgear Room 3.OE-3

Thirteen generic cases were examined on the basis of

these assumptions and conditions. They represent those

cases presented in Table 4.2. Details of calculations are

presented in Appendix E of Reference 9. Calculations for

core damage frequency, risk, and sensitivity studies were

performed using the Top Event Matrix Analysis Code

(TEMAC) (Ref. 14) and the Latin Hypercube Sampling

Code (Ref. 15).

NUREG-1472 6
Table 4.4 FPS Actuation Frequency per System-Year

Hardware

Human Error Failure Other Total

Water Preaction 5.2E-4 5.2E-4 5.2E-4 1.6E-3

Water Wet Pipe 2.8E-3 5.OE-3 1.8E-3 9.6E-3

Water Deluge 9.4E-3 5.5E-3 1.3E-2 2.8E-2

Halon 3.5E-3 5.3E-3 8.8E-4 9.7E-4

CO 2 1.4E-3 2.3E-3 1.8E-3 5.5E-3

4.1.3 Generic Core Damage Frequency

Estimates

This section examines core damage frequency (CDF) in

three generic plant cases. The following caveats must be

kept in mind when reviewing the data presented in these

cases:

- The generic plants examined are represented by only

three fire zones: the cable spreading room, the diesel

generator rooms, and the emergency electrical

switchgear rooms. In any given specific plant, other

fire zones may be significant or even dominant contributors to CDF, and specific plant analysis must be

conducted to evaluate an individual plant (this was

one of the reasons for initiation of the IPEEE program).

- Because the level of damage to electrical cables and

electro-mechanical components that results from

short term exposure to Halon FPS agent is not clear

from experimental and historical data, Halon is not

considered in evaluation of the best case plant. If the

assumption is made that Halon presents no short

term threat to cables and components, then the incremental CDF associated with a generic plant with

all Halon FPS systems would be only that resulting

from Root Cause 12 (seismic/fire interaction).

Therefore, the CDF associated with Halon suppressant agent damage to cables and components would

be zero.

- CDF data in the tables representing the generic

cases are mean values for the individual root cause

and room totals, and the sum of mean values for the

overall plant values. Uncertainty calculations were

not accomplished in examining the overall generic

plant CDF values.

- In the typical and best generic plant cases, while in

the CDF tables a value is presented for the emergency electrical switchgear rooms, it must be recognized

that data (Appendix D to Reference 9) indicate that

an automatic FPS is installed in these rooms in only

about 20 percent of the U.S. commercial nuclear

power plants. In the remaining 80 percent of these

plants, there is no automatic FPS system installed.

Thus for these generic cases, two overall values for

CDF are provided, one for the case with an FPS in

the emergency electrical switchgear rooms, and one

for the case with no automatic FPS in those rooms. It

is suspected that although Generic Issue 57 associated CDF for the case with no automatic FPS installed is lower, the overall plant CDF may not be

lower because of the likelihood that the CDF associated with fire in these rooms is higher without an

automatic FPS systems. A detailed analysis of this

issue was not conducted.

- For all the cases, the LLNL seismic hazard curves were

used.

4.1.3.1 Most Vulnerable Generic Plant

In the most vulnerable case plant, for each of the three

rooms examined (cable spreading, diesel generator, and

emergency electrical switchgear rooms) the FPS system

resulting in the highest CDF was selected. In all cases, the

resulting FPS system is deluge water. Additionally, for

this case only, mercury wetted contact type relays were

assumed to be installed in the deluge FPS control system.

Data for this case are presented in Table 4.5. For the worst

case generic plant, CDF is calculated to be 2.5E-3/year [of

which 2.2E-3/year (88%) is due to the deluge water system in the diesel generator rooms]. This CDF is sufficiently high to require consideration of plant specific corrective actions on any plant whose configuration actually

contains most or all of the assumed features of the hypothetical "most vulnerable" plant.

This hypothetical "most vulnerable" generic plant is assumed to be subject to the seismic hazard curve for the

ANO site, including the recent New Madrid seismic

event, and it is assumed that any seismic event from 1/2

SSE to 6 SSE will cause actuation of the plant's FPS (due

to assumptions "a" and "b", described below for the FPS's

configuration). This high CDF results from the FPS configuration assumptions that the plant either: a) has seismically sensitive relays in the FPS, orb) is actuated solely by

a smoke sensitive detector; and c) that vulnerable components of the diesel generator are susceptible to water

intrusion from the FPS.

7 NUREG-1472
Table 4.5 Core Damage Frequency - "Most Vulnerable" Generic Plant

Fire Zone: Cable Spreading Diesel Generator Emergency Electrical

Room Rooms Switchgear Rooms

FPS Type: Deluge Water Deluge Water Deluge Water

Root Cause

4 1.6E-6 1.8E-6

5 2.6E-8 3.1E-6

6 9.1E-7 1.1E-6

7 2.6E-5

8 2.6E-5

9

7/8 2.2E-3 2.2E-4

8/9

7/8/9

11 4.4E-7

12 1.2E-5 1.2E-5

13 2.3E-6 2.5E-6

Total 6.9E-5 2.2E-3 2.4E-4

Total CDF for "Most Vulnerable" Generic Plant: 2.5E-3/RY.

Efforts are underway to identify any actual plants with

deluge water systems in the diesel generator rooms whose

configuration meets assumptions (a) or (b) and (c); however, based on a preliminary survey, it is expected that very

few (if any) will be identified in this category. Consideration of plant-specific followup will be given on a case-bycase basis.

4.1.3.2 Typical Case Generic Plant

In the typical case plant, for each of the three rooms

examined, the FPS was chosen that represents the most

common installation in commercial U.S nuclear power

plants. Based on survey data (Appendix D to NUREG/

CR-5580), there are 69 cable spreading rooms with wetpipe water FPS systems, 49 diesel generator rooms with

CO 2 FPS systems, and 18 switchgear rooms with CO2 FPS

systems. CDF data associated with these systems in the

generic plant are shown in Table 4.6. For the typical case

generic plant, CDF is calculated to be 3.5E-5/year with an

automatic FPS in the emergency electrical switchgear

rooms, and 2.4E-5/year with no automatic systems installed in the emergency electrical switchgear rooms.

4.1.3.3 Least Vulnerable Generic Plant

In the least vulnerable case plant several assumptions are

made'to optimize the plant for minimum CDF associated

with Generic Issue 57. The assumptions are based on the

information gained from the study. The assumptions are:

- For each of the three rooms examined, the FPS

system resulting in the lowest CDF was selected.

Accordingly, a CO2 FPS was selected for the cable

spreading room, a preaction water FPS was selected

for the diesel generator rooms, and a CO2 FPS was

selected for the emergency electrical switchgear

rooms.

For the cable spreading room, it was assumed that

there are no electrical cabinets in the room to act as

fire sources in a seismic event. This assumption is

consistent with some, but not all of the individual

plants walked-down. This eliminates CDF associated with Root Cause 12 (seismic/fire interaction)

in this space.

For the emergency electrical switchgear rooms, the

electrical cabinets that remain energized in a LOSP

event are assumed to be seismically restrained

against sliding or tipping, eliminating the CDF associated with Root Cause 12 (seismic/fire interaction).

This kind of cabinet restraint was observed in some,

but not all of the plants walked-down.

For all three fire zones, relays in the FPS control

systems are assumed to be seismically qualified, and

the CDF associated with relay chatter is reduced by a

factor of 10 from that in the typical generic plant

case. Such seismically qualified relays were found in

some, but not all, of the plants walked-down.

Data for this case are shown in Table 4.7. It should be

noted that to achieve further reductions in CDF, the contributions due to non-seismic root causes must be reduced. The principal factors involved that must be reduced are the conditional probabilities for damage of

cables and active electro-mechanical components, given

that they are wetted by a fire suppressant agent. In this

NUREG-1472 8
Table 4.6 Core Damage Frequency - "Typical" Generic Plant

Fire Zone: Cable Spreading Diesel Generator Emergency Electrical

Room Rooms Switchgear Rooms

FPS Type: Wetpipe Water CO2 CO2

Root Cause

4 4.7E-8 1.5E-8

5 2.6E-7

6 8.4E-8 2.5E-8

7

8 1.2E-5

9

7/8 1.6E-6

8/9

7/8/9

11

12 1.2E-5 9.4E-6

13 2.9E-8 2.1E-8

Total 1.2E-5 1.2E-5 1.1E-5

Total CDF for "I1,pical" Generic Plant: 3.5E-5/year with an automatic FPS installed in the emergency electrical

switchgear room, 2.4E-5/RY without.

Table 4.7 Core Damage Frequency - "Least Vulnerable" Generic Plant

Fire Zone: Cable Spreading Diesel Generator Emergency Electrical

Room Rooms Switchgear Rooms

FPS Type: CO 2 Preaction Water CO2

Root Cause

4 2.3E-7 1.5E-8

5 2.6E-8 2.6E-7

6 3.9E-7 2.5E-8

7

8

9

7/8 1.2E-7 1.6E-7

8/9

7/8/9 < 1.OE-8

11

12 < 1.0E-8 < 1.OE-8

13 3.OE-7 2.1E-8

Total 1.1E-6 < 1.0E-8 4.8E-7

Total CDF for "Least Vulnerable" Generic Plant: 1.6E-6/RY with an automatic FPS installed in the emergency

electrical switchgear room, 1.1E-6/RY without.

9 NUREG-1472
study, some of these values had to be established using

zero data point bounding methods, while the remainder

are based on very few documented actual damage events.

A testing program could better define these conditional

probabilities, and in all likelihood result in reduced calculated values for non-seismic root cause contributions to

CDR

For the "least vulnerable" case generic plant, CDF is

calculated to be 1.6E-6/RY with an automatic FPS in the

emergency electrical switchgear rooms, and 1.1E-6/RY

with no automatic systems installed in the emergency

electrical switchgear rooms.

Table 4.8 summarizes the CDF contributions from each

root cause for the three plant-specific evaluations.

4.2 Dose Consequences Analysis

For purposes of this study, consequences are measured in

person-sievert, abbreviated as person-Sv, with equivalent

consequences in units of person-rem given enclosed in

parentheses immediately following. Also, benefits are

given in person-Sv (person-rem) averted. Once the CDF

and changes in CDF from a potential resolution alternative have been calculated (Section 4.1), the next step is to

calculate the corresponding consequences in person-Sv

(person-rem), and hence, benefits in person-Sv (personrem) averted.

Section 4.7 of Reference 9 contains a detailed discussion

of the generic offsite dose calculations and presents generic risk values for each of the applicable root causes.

In addition to the plant-specific evaluations performed,

three generic plant cases were analyzed representing

design configurations and fire protection system type utilization in key plant locations so that "most vulnerable,"

"typical," and "least vulnerable" cases were identified.

These assumed plant configurations may not necessarily

correspond to the specific design of a given plant, but

rather they represent a wide spectrum of possible designs,

with the "typical" case judged to be the closest to the

design and associated risk level of the predominant number of plants. The basis for this judgement is presented

later in this section.

The generic plants examined are represented by only

three fire zones: the cable spreading room, the diesel

generator rooms, and the emergency electrical switchgear

rooms. In any given specific plant, other fire zones may be

significant or even dominant contributors to CDF and

risk,.and IPEEE plant specific analysis must be conducted

to evaluate an individual plant.

Because the level of damage to electrical cables and

electro-mechanical components that results from short

term exposures to Halon FPS agent is not clear from

experimental and historical data, Halon FPS systems are

not considered in evaluation of the "least vulnerable

case" plant. If the assumption is made that Halon presents

no short term threat to cables and components, then the

incremental CDF and risk associated with a generic plant

with all Halon FPS systems would be only that resulting

from Root Cause 12 (seismic/fire interaction). CDF and

risk associated with Halon suppressant agent damage to

cables and components would be zero.

Table 4.8 Plant-Specific Base Case Results in Terms of Core Damage Frequency

(Per Reactor Year)

Westinghouse

Root Cause Plant B&W Plant GE Plant

1. N/A N/A 5.7E-7

2. N/A N/A N/A

3. N/A N/A N/A

4. 1.4E-6 2.3E-6 3.3E-7

5. 1.1E-6 N/A 2.3E-8

6. 2.1E-6 1.4E-6 5.4E-7

7. N/A N/A 3.3E-7

8. 2.6E-7 1.5E-6 1.1E-5

9. N/A < 1.OE-8 N/A

10. N/A N/A 6.9E-7

11. 4.2E-7 6.4E-7 5.7E-7

12. 1.4E-6 4.7E-5 8.6E-6

13. 5.7E-7 2.9E-6 4.4E-7

Total 7.3E-6 5.6E-5 2.3E-5

NUREG-1472 10
CDF and risk data in the tables representing the generic

cases are mean values for the individual root cause and

room totals, and the sum of mean values for the overall

plant values. Uncertainty calculations were not performed in examining the overall generic plant CDF and

risk values. However, uncertainty could be expected to be

distributed in a way similar to that in the specific plant

analysis. From 5% to the 95% point in composite CDF,

the range was about two orders of magnitude. For risk, the

range would be expected to be considerably larger.

For purposes of this regulatory analysis the results of the

evaluation of the "typical" plant case were used to assess

the cost/benefit parameters associated with this issue.

In the typical case plant, for each of the three rooms

examined, the FPS was chosen that represents the most

common installation in commercial U.S. nuclear power

plants. Based on survey data (Appendix D to Reference 9), there are 69 cable spreading rooms with wetpipe

water FPS systems, 49 diesel generator rooms with C02

FPS systems, and 18 switchgear rooms with C02 FPS systems. Dose consequences data (PWR and BWR) associated with these systems in the generic plant are shown in

Thbles 4.9 and 4.10. For the typical case PWR, 20 year risk

is 0.54 personýSv (54 person-rem) with an automatic FPS

installed in the emergency electrical switchgear room,

and 0.51 person-Sv (51 person-rem) without. For the typical case BWR, 20 year risk 2.2 person-Sv (220 person-rem)

with an automatic FPS installed in the emergency electrical switchgear room, and 2.1 person-Sv (210 person-rem)

without.

4.3 Cost Analysis

To calculate the cost for the various backfit alternatives

the analysis drew from several sources and followed the

guidelines of References 16 and 17. The computer code

FORECAST 3.0 (Ref. 18) which incorporates the information in the preceding references was used to develop

estimates for the various backfit alternatives. For each

alternative the costs noted in Section 4.3.1 through 4.3.9

were considered.

4.3.1 Labor and Equipment/Materials Costs

The Energy Economic Data Base (EEDB), Reference 19

(embedded in the FORECAST code), and R.S. Means

Cost Guides (Ref. 20) provided the basis for the equipment/material cost and labor estimates. The EEDB incorporates "as-built" cost information (both material unit

cost and installation labor hours) for nuclear plant construction activities. The material and labor information

from R.S. Means Cost required adjustment to the specified EEDB basis to properly reflect the nuclear plant level

of effort and equipment/material specifications. Two factors, derived for and used in previous cost study (Ref. 21)

were employed: Means-EEDB equipment/materials

costs were adjusted by multiplying by 2.1 and MeansEEDB labor hours were adjusted by multiplying by 2.7.

The cost modification factors of 2.1 for the equipment/

material costs and 2.7 for the labor costs were utilized to

Table 4.9 Risk in person-Sv/RY (person-rem/RY) "Typical" PWR Plant

Fire Zone: Cable Spreading Diesel Generator Emergency Electrical

Room Rooms Switchgear Rooms

FPS Type: Wetpipe Water CO 2 CO2

Root Cause

4 3.6E-5 (3.6E-3) 1.1E-5 (1.1E-3)

5 2.OE-4 (2.OE-2)

6 6.4E-5 (6.4E-3) 1.9E-5 (1.9E-3)

7

8 9.1E-3 (9.1E-1)

9 4.OE-5 (4.OE-3)

7/8 1.2E-3 (1.2E-1)

8/9

7/8/9

11

12 9.1E-3 (9.1E-1) 7.2E-3 (7.2E-1)

13 2.2E-5 (2.2E-3) 1.6E-5 (1.6E-3)

Total 9.3E-3 (9.3E-1) 9.1E-3 (9.1E-1) 8.6E-3 (8.6E-1)

Total risk for "Typical" PWR Generic Plant: 0.027 person-Sv/RY (2.7 person-rem/RY) with an automatic FPS installed in

the emergency electrical switchgear room, 0.026 person-Sv/RY (2.6 person-rem/RY) without.

11 NUREG-1472
Table 4.10 Risk in person-Sv/RY (person-rem/RY) "Typical" BWR Plant

Fire Zone: Cable Spreading Diesel Generator Emergency Electrical

Room Rooms Switchgear Rooms

FPS Type: Wetpipe Water CO 2 CO2

Root Cause

4 1.5E-4 (1.5E-2) 4.6E-5 (4.6E-3)

5 7.9E-4 (7.9E-2)

6 2.7E-4 (2.7E-2) 7.6E-5 (7.6E-3)

7

8 3.7E-2 (3.7E + 0)

9 1.7E-4 (1.7E-2)

7/8 4.9E-3 (4.9E-1)

8/9

7/8/9

11

12 3.9E-2 (3.9E + 0) 2.9E-2 (2.9E + 0)

13 9.4E-5 (9.4E-3) 6.4E-5 (6.4E-3)

Total 3.9E-2 (3.9E + 0): 3.7E-2 (3.7E + 0) 3.5E-2 (3.5E + 0)

Total risk for "lypical" BWR Generic Plant: 0.11 person-Sv/RY (11 person-rem/RY) with an automatic FPS installed in

the emergency electrical switchgear room, 0.11 person-SvIRY (11 person-rem/RY) without.

reflect costs incurred at nuclear power plants. The R.S.

Means Cost Guide reflects costs for non-nuclear facilities

and the costs must be modified to represent the level of

effort and material/equipment specifications required at

a nuclear power plant.

Additionally, for operating nuclear power plants there are

a number of workplace characteristics which significantly

reduce the level of productivity and thus increase the

number of labor hours required to accomplish a task.

These characteristics, discussed in detail in FORECAST

3.0, include access, congestion and interference, radiation, and task management. Since EEDB reflects only

new (or "as-built") plant conditions, the installation labor

hours were adjusted to properly consider actual conditions existing at operating nuclear plants.

The total labor costs associated with the proposed modifications include overhead charges (at 100 percent of direct

labor) to account for contractor management, administrative support, rent, insurance, etc.

4.3.2 Engineering and Quality Assurance/

Control Costs

These costs reflect the cost of engineering and design, as

well as quality assurance/control (QA/QC) activities associated with implementing the requirements. For requirements affecting structures/systems already in-place (operating plants) the guidelines of abstract 6.4 of "Generic

Cost Estimates," (Ref. 17) recommend that a 25 percent

engineering and QAIQC factor be applied to the diect

cost (i.e., labor and materials and cost but without any

overhead charges). All cost estimates developed in this

study included this engineering and QA/QC cost component.

4.3.3 Radiation Exposure

Worker radiation exposure estimates were derived based

on guidelines presented in Abstract 5.1 of Reference 17.

The collective radiation exposure associated with the implementation of a proposed plant modification is estimated by taking the product of the in-field labor hours

necessary to perform the task and the work area dose rate

associated with that particular task.

In this study the work area in which the modifications

would take place are considered to be either low-dose

contaminated areas (cable vault/tunnel) or clean areas

(diesel generator rooms). Therefore, radiation exposure

is either minimal or zero for the modifications proposed in

this study.

4.3.4 Health Physics Support Costs

Health physics requirements for the potential plant enhancements were developed based on information and

guidelines presented in Abstract 2.1.6 of Reference 17.

Two factors were considered; the size of the work crew

and the magnitude of the radiation field. The plant health

physics (HP) personnel perform radiation surveys that are

conducted throughout the time required to perform the

modification, staff radiological checkpoints, set up anticontamination clothing removal areas, as well as determine badging requirements.

NUREG-1472 12
Some of the modifications are performed in low radiation

but contaminated work areas, such as the cable vault/tunnel. Therefore, the health physics support costs are highest for this type of improvement. However, a minimum

health physics cost increment is associated even with

physical modifications conducted in clean work areas

since area radiological surveys and other HP activities still

have to be performed.

4.3.5 Anti-Contamination Clothing Costs

Cost estimates for anti-contamination (anti-c) clothing

used while performing the plant modifications were

derived based on Abstract 2.1.5 of Reference 17. The cost

per suitup assumes that each member of the work crew

requires two complete sets of anti-c clothing per eighthour shift. Included in the cost per suit-up area the cost of

purchasing the anti-c clothing set, its wear-out rate, laundering costs, etc. Only work tasks conducted in contaminated plant areas were considered to include this cost

increment.

4.3.6 Radioactive Waste Disposal Cost

The costs for disposal of radioactive wastes generated

during plant modifications were derived based on guidelines of Abstract 2.1.4 of References 17. For the study the

cost increment associated with the disposal of radioactive

wastes is applicable only to those plant modifications that

necessitated removal of existing system components located in a contaminated area. The costs are, however,

insignificant (less than five percent of total cost).

4.3.7 Other Licensee Costs

Other costs incurred by the utility as a result of implementing the proposed physical plant modifications included the costs of re-writing procedures, training the

staff (both maintenance and operations), and changing

recordkeeping or reporting requirements. For each of the

above stated cost categories, the costs were derived following the guidelines presented in Abstracts 2.2.2, 2.2.3,

and 2.2.4, respectively, of Reference 17. In this study, for

some of the plant modifications proposed, these costs

represent a significant portion of the total cost.

4.3.8 NRC Costs

These cost represent NRC implementation costs. They

account for such NRC activities as developing inspection

guidelines and procedures, assuring compliance with the

proposed regulatory action, and other technical tasks. In

this study, the cost estimates associated with the NRC

were primarily derived from guidelines and input provided by References 16 and 18.

4.3.9 Onsite Averted Costs

In addition to the costs associated with the modifications,

an evaluation of the costs associated with the potential

reduction of severe onsite consequences were evaluated.

"A Handbook for Value-Impact Assessment" (Ref. 16)

was used as the reference for performing this evaluation.

The values for onsite averted cost were calculated using

the following equation:

Vop = NU (F0 -_Fn )

Where

Vop = the cost of avoided onsite property damage

N = the number of affected facilities

U = the present value of onsite property damage given a release

F0 = the original core damage frequency (base

case)

Fn = the core damage frequency after implementing an option

and

c . e-rti t

U I= (-r2) •(1 - e r(tf _ ) (1-er m )

where

c = cleanup, repair, and replacement power

costs

tf = years remaining until end of plant life

ti = years before reactor begins operation

m = period of time over which damage costs

are paid out

r = discount rate (for 10%, r = 0.10)

The cost handbook (Ref. 16) recommends best estimate

values for input to calculating U (present value of onsite

property damage given a release) as follows:

c = $1,650x 106

m = 10 reactor-years

r = 0.10

tf = 20 reactor-year

ti = 0 reactor-years

Using the above values for calculating U yields the following result:

Best estimate - $9.0 billion/severe accident event

This value is then applied to the potential change in accident frequency, or these analyses, change in core damage

frequencies for each option.

13 NUREG-1472
4.4 Cost Estimate Uncertainties

The areas of uncertainty associated with the cost estimating model for this study included the following:

1. Labor rate variations due to plant site location

2. Contingency allowance

3. Variability of in-plant work environment conditions

4. Licensee procedural/administrative/analytical conditions

5. NRC procedural/administrative/analytical cost

6. Discount rate variation in the recurring cost module

7. Waste disposal cost module

Each cost estimate was evaluated to determine all areas of

uncertainty applicable. Specific numerical values were

used for each individual cost analysis. More detailed discussions of cost uncertainties may be found in Section

5.1.6 of Reference 9.

4.5 Backfit Alternatives Cost

Estimates

A discussion of the proposed backfit alternatives based on

plant walkdowns and evaluations is presented in the following subsections.

4.5.1 Alternative 1 - No Action

Under this alternative resolution there will be no cost

involved.

4.5.2 Alternative 2 - Upgrade a FPS

Actuation Controller with Seismically

Qualified Printed Circuit Boards

Because of concerns and industry experience with relay

chatter during seismic events it may be prudent to investigate replacing existing relays in the FPS controller cabinets with printed circuit boards. Based on plant specific

walkdowns many types of FPS actuation relays were

found. For some plants, mercury wetted relays for FPS

actuation and/or the annunciation of alarms and isolation

of room cooling was found. Given a seismic event, there is

a high likelihood of actuation for some of these relay

types. The intent of this modification is to replace the

relays with printed circuit boards and prevent any damage

which may result from inadvertent FPS actuation during a

seismic event. For each of the areas modified, the contribution to the core damage frequency from Root cause 8

(relay chatter in a seismic event) would be eliminated.

For the purpose of this estimate, it is assumed that this

modification could be performed during a planned unit

outage. Therefore, costs associated with unit shutdown or

startup and replacement power costs are not included in

the estimate. It is assumed that this type of activity has

been done many times before, requiring no learning curve

adjustments. No significant radwaste disposal is involved.

Also, the costs for security and fire watch personnel are

estimated.

The total cost to implement this plant modification ranges

from $13,000 to $17,000.

4.5.3 Alternative 3 - Replace Smoke Detector

Actuated FPS with Heat Detector

Actuated System

Replacing an existing smoke detector actuated system.

with a heat detector actuated system will eliminate contributions from Root Cause 1, 7 and 10 which are smoke

detector specific. However, to provide an additional detection capability, it may be prudent to leave the existing

smoke detectors intact for indication purposes only.

As with the upgrade of FPS controllers in the previous

section, it is assumed that this project could be completed

during a scheduled outage. No costs associated with shutdown, startup, and replacement power are included. A

general set of productivity factors representative of a

cable spreading room is used. Cost for security personnel

and a fire watch (306 person-hours) are included in the

total installation labor cost.

The cost to implement this option would range from

$78,000 to $105,000.

4.5.4 Alternative 4 - Reroute Safety-Related

Cables

Through plant walkdowns it had been found that there are

certain "pinch points" located in the plant where cabling

for certain redundant safety-related systems are run together. Given a fire or FPS actuation that could damage

these cables, these safety systems are vulnerable to simultaneous failure. Therefore the intent of this modification

is to reroute one of the sets of cabling to remove it from a

common failure vulnerability.

For the purposes of this estimate, it is assumed that the

old cable run would be eliminated in place and that the

new cable installation could be completed during a

planned unit outage. Therefore costs associated with unit

shutdown or startup and replacement power costs are not

included in the estimate. It is further assumed that one

length of cable would be required (only control) and that

the cable would need to be qualified for harsh environments. Thus, an estimate is made for cable subject to the

requirements of the plant equipment qualification proNUREG-1472 14
gram. Costs for cable in conduit are used for all of the

cable runs. The total length of cable run in conduit required is assumed to be approximately 152 meters (500

feet). Ten penetrations are required, and terminations are

needed at both ends of the cable run.

Separate environmental and labor productivity factors are

used for the two main plant areas. It is assumed that this

type of activity has been done many times before, requiring no learning curve adjustments. It is assumed that part

of the rerouting will be done in a radiation area and as

such appropriate factors and costs for anti-c clothing and

HP support are included. No significant radwaste disposal

is involved. Also, the costs for security and fire watch

personnel are estimated.

The total cost to implement this plant modification ranges

from $136,000 to $185,000.

4.5.5 Alternative 5 - Seismically Qualify the

CO2 Tank, Outlet Piping, and Battery

Rack

A seismic walkdown and subsequent fragility analysis for a

typical C02 system found a high likelihood of suppressant

agent diversion given an earthquake. Failure, of the tank

or its outlet piping or the FPS battery dominated the

overall probability of failure of the C02 system. This potential plant modification would seismically qualify the

C02 tank, battery, and its immediate outlet piping.

For the modification, it is assumed that this project would

not require a special plant shutdown to implement.

Therefore, shutdown, startup and replacement power

costs are not included. Given that there is no real potential for contamination, cost estimates for health physics

support and anti-c clothing are not included. Labor hours

were included for security and fire watch support.

The total cost for this upgrade ranges from $97,000 to

$131,000.

4.5.6 Alternative 6 - Seismically Qualify A

FPS Battery Rack

It was found during a plant specific walkdown, that a water

based FPS for an entire plant had two electric driven fire

pumps and one diesel driven fire pump. Given a LOSP,

both electric driven pumps would fail due to being powered from non-vital busses. A failure of the starting battery for the diesel pump would lead to a loss of a fire main

pressure.

For this modification it is assumed that this project would

not require a special plant shutdown to implement.

Therefore, shutdown, startup and replacement power

costs are not included. Given that there is not a real

potential for contamination, cost estimates for health

physics support and anti-c clothing are not included. Labor hours are included for security and fire watch support.

The total cost for this ranges from $35,000 to $42,000.

4.5.7 Alternative 7 - Upgrade the FPS Water

Quality

Water is the most frequently used fire suppressant at

nuclear power plants. The most common delivery systems

are pre-action sprinklers, deluge wet-pipe sprinklers, drypipe sprinklers and standpipe hose systems. It has been

postulated that improvement of water quality will reduce

the potential for damage to safety-related components

from exposure to water suppressant. This is based on the

thought that pure water, less conductive than normal fire

fighting water, would be less likely to cause short circuits

or grounds. To address this concern, a modification to

upgrade the FPS water quality is performed. For this

modification a water purification system would be required along with the associated piping and valves and a

storage tank. It is assumed for this modification that the

existing FPS piping and pumps would be utilized. Also, it

is assumed that this project could be performed during

plant operation. Therefore, shutdown, startup and replacement power costs are not included. Given that there

is not a real potential for contamination, costs estimates

for health physics support and anti-c clothing are not

included. Labor are hours included for security and fire

watch support.

The total cost for this upgrade ranges from $1,174,000 to

$1,577,000.

4.5.8 Alternative 8 - Replace Deluge with

Preaction Sprinkler FPS

Of all the types of water based FPS utilized in nuclear

power plants the pre-action sprinkler system has been

found to experience the least number of inadvertent actuations (Ref. 9). A pre-action sprinkler system requires

the opening of a deluge valve, either automatically, with a

control signal, or manually, and a rise in the ambient

temperature to the melting point of the fusible links on

the sealed sprinkler heads. A deluge FPS only requires

the opening of a valve, either automatically or manually to

discharge water under pressure to the open spray heads.

This modification involves replacing a deluge FPS with a

pre-action sprinkler FPS. This modification would reduce

the frequency of inadvertent actuations and localize the

application of FPS agent.

For this modification fusible link sealed sprinkler heads

would need to be added to the existing deluge FPS. All of

the existing hardware would be kept in place. It is also

assumed the existing locations of sprinkler heads would be

adequate for preaction system. This modification will not

require a special plant shutdown to implement. There15 NUREG-1472
fore, shutdown, startup and replacement power costs are

not included. Also this modification will not require any

health physics support or anti-c clothing. Labor hours are

included for security and fire watch support.

The total cost for this upgrade ranges from $22,000 to

$30,000.

4.5.9 Alternative 9 - Replace Electrical

Cabinet with a Cabinet Designed to

Prevent Water Intrusion

One study (Ref. 11) has reported that electrical equipment failure modes related to water were mainly due to

electrical shorting and long term corrosion. Wherever

water intrusion is possible the equipment could be expected to fail through shorting, grounding, tripping of

overcurrent devices, physical damage due to the velocity

of direct hose streams, or long term corrosion causing

potential failure of electro-mechanical parts. These failure modes were found to be dependent on the national

electrical manufacturers associated (NEMA) rating or

the configuration of the electrical enclosures. The appropriate NEMA rating to preclude water intrusion can potentially eliminate failures in electrical equipment due to

water based FPSs. Enclosures that have a NEMA rating

of 1 and 5 are subject to water intrusion under all water

spray conditions applicable for this study. Enclosure with

NEMA ratings of 2, 3, 3R, 3S, 4, 4S, 6, 6P, 11, 12, 12K and

13 are expected to prevent water intrusion under direct

hose stream and splashing water spray. Only those enclosures with a NEMA rating of 6 and 6P are expected to

prevent water intrusion under temporary submersion due

to flooding. The intent of this modification is to replace

existing safety-related electrical cabinets and enclosures

with NEMA spray-proof rated enclosures. Although the

modified electrical cabinets may require internal cooling

a cost for internal cooling was not included for this analysis.

Unlike the other generic modifications presented in this

section this modification can not be completed during a

scheduled outage since safety-related cabinets would

need to be de-energized. Therefore, costs associated with

shutdown, startup and replacement power are included.

Additionally, costs for health physics support and anti-c

clothing are included. Costs for security and fire watch are

also included. The cost to implement this modification

would range from $22,000 to $30,000 per cabinet.

4.5.10 Alternative 10 - Seismically Anchor

Safety-Related Cabinets Susceptible to

Tipping/Sliding Failure

In a seismic event energized cabinets may present a potential source for fire. Although it is assumed that in a

seismic event offsite power will be lost, thus deenergizing

many electrical cabinets, there will be a number of safetyrelated cabinets energized by alternative power sources

(batteries, diesel generators). These energized cabinets

are susceptible to tipping/sliding failure possible leading

to a fire.

It is assumed that this modification would eliminate the

potential for seismically induced fires due to the tipping/

sliding failure of an energized electrical cabinet. For this

modification a plant shutdown will not be required.

Therefore, shutdown, startup and replacement power

costs are not included. Additionally, it is assumed there is

no real potential contamination so that cost estimates for

health physics support and anti-c clothing are not included. Labor hours are included for security watch support. The total cost for this modification ranges from

$67,000 to $91,000.

NUREG-1472 16
5 VALUE/IMPACT ANALYSIS

The value/impact (V/I). methodology for analyzing the

various alternatives examined under this study is based on

the requirements of the backfit rule (10 CFR 50.109) and

related implementing guidance contained in References

16, 22, and 23. One of the primary considerations here is

the derivation of cost/benefit ratios for each alternative

evaluated in terms of cost in dollars per person-Sv (dollars

per person-rem) averted, which may be compared to a

guideline such as $100,000 per person-Sv ($1,000 per person-rem). This quantitative guidance is one of the elements considered in the decision-making process. Also,

the interim guidance contained in Reference 10 is consistent with the objectives set forth in Section 2. Deterministic considerations on the merits of a proposed alternative

resolution are also part of the decision with respect to a

given alternative (Section 6). In the following subsections

a description of each alternative and the results of a value/

impact assessment are presented for backfit as well as

frontfit cases.

5.1 Backfit Analysis

Potential backfits, were identified as a result of plant

walkdowns and evaluations and .those analyzed are presented here as alternatives to resolve this issueý

5.1.1 Alternative 1 - No Action

Under this alternative there would be no new regulatory

requirements. Consistent with existing regulations, this

alternative does not preclude a licensee, or an applicant

for a license, from proposing to the NRC staff design

changes intended to enhance the reliability or operability

of the fire protection systems and their components on a

plant-specific basis.

5.1.2 Alternative 2 - Upgrade an FPS

Actuation Controller with Seismically

Qualified Printed Circuit Boards

A cost for this modification of $14K (one FPS controller)

was estimated. This modification is assumed to eliminate

any contribution of risk from Root Cause 8 or any combination of Root Cause 8 scenarios. It is assumed that for

the cable spreading room (CSR) one FPS controller will

be replaced and for the emergency diesel generator

(EDG) and emergency switchgear (ESGR) areas two FPS

controllers will be replaced.

Table 5.1 presents the cost/benefit results for this modification. Even without OSAC included this option appears

to be cost-beneficial for EDG areas with a deluge or C02

system and for the ESGR with a deluge FPS. When

OSAC are included a Halon FPS in the CSR for both a

PWR and BWR and a deluge and C02 FPS in the CSR for

BWR appears to be cost-beneficial.

5.1.3 Alternative 3 - Replace Smoke Detector

Actuated FPS with a Heat Detector

Actuated System

A cost for this modification of $87K was estimated. This

modification was proposed only for the cable spreading

room and would eliminate any incremental contribution

from Root Cause 7 to core damage frequency from this

area. The risk reduction for this modification is 0.45 person-Sv (45 person-rem) for the PWR analyzed and 1.0

person-Sv (180 person-rem) for the BWR analyzed. Table

5.2 presents the cost/benefit results for this modification.

If OSAC are not included, this modification appears to be

beneficial only for the BWR analyzed. If OSAC are included both the PWR and BWR analyzed indicate a beneficial option.

5.1.4 Alternative 4 - Reroute Safety-Related

Cables

A cost for this modificationof $154K was estimated. This

modification is assumed to apply only to the cable spreading room and was assumed to reduce risk in the CSR by an

order of magnitude. If OSAC are not included this modification Appears to only be beneficial for a deluge FPS in a

BWR. With OSAC included, a deluge FPS in both a PWR

and a BWR and a wetpipe and preaction FPS in a BWR

indicate a beneficial modification. Table 5.3 presents the

cost/benefit results for this modification.

5.1.5 Alternative 5 - Seismically Qualify the

CO 2 Tank, Outlet Piping, and Battery

Rack

For this modification it 'was assumed that there is one

common CO 2 tank per plant. A cost for this modification

of $109K was estimated. Implementing this modification

eliminates the incremental contribution to core damage

frequency from Root Cause 12 (seismic/fire interaction)

for CO2 systems. When OSAC are included for the cable

spreading room and emergency switchgear room for the

BWR analyzed, a beneficial modification is indicated.

Table 5.4 presents the cost/benefit results for this modification.

17 NUREG-1472
Table 5.1

Alternative 2

$K/person-Sv ($K/person-rem) averted

Preaction Deluge Wetpipe CO 2 Halon

PWR

CSR

EDG

ESGR

CSR

EDG

ESGR

BWR

CSR

EDG

ESGR

CSR

EDG

ESGR

6.7E + 04

(6.7E + 02)

2.5E + 05

(2.5E + 03)

6.6E + 04)

(6.6E + 02

2.5E + 05

(2.5E + 03)

7.8E + 02

(7.8E + 00)

1.9E + 01

(1.9E-01)

LOE + 02

(1.0E+00)

2.3E + 02

(2.3E + 00)

< .OE + 02

(< .OE + 00)

<l.OE+02

(< .OE + 00)

1.9E + 02

(1.9E + 00)

4.9E + 00

(4.9E-02)

2.5E +02

(2.5E + 00)

5.5E + 01

(5.5E-01)

< 1.OE + 02

(< .OE + 00)

< .OE + 02

(< .OE + 00)

7.OE + 02

(7.OE + 00)

9.OE + 01

(9.OE-01)

1.1E+03

(1.lE+01)

1.6E + 02

(1.6E + 00)

< 1.OE+02

(< .OE + 00)

5.2E + 02

(5.2E + 00)

1.8E + 02 1

(1.8E +00)

2.3E+ 01

(2.3E-01)

5.6E + 02

(5.6E + 00)

4.E + 01

(4.OE-01)

<1.OE + 02

( < 1.OE + 00)

1.2E+02

(1.2E + 00)

6.4E + 02

(6.4E + 00)

2.2E+03

(2.2E + 01)

2.2E + 03

(2.2E + 01)

1.OE + 02

(1.OE + 00)

1.6E + 03

(1.6E + 01)

1.6E + 02

(1.6E + 00)

2.6E + 01

(2.6E-01)

4.2E + 02

(4.2E + 00)

1.6E + 04

(1.6E + 02)

637E +04

(6.7E +02)

M.E +01

(2.SE-01)

1 .6E + 04

(1.6E + 02)

6.7E + 04

(6.7E + 02)

Bold type indicates OSAC included

5.1.6 Alternative 6 - Seismically Qualify A

FPS Battery Rack

The cost for this modification was estimated to be $39K.

This modification is assumed to eliminate Root Cause 12

core damage frequency contributions from water-based

FPSs. Table 5.5 presents the cost/benefit results for this

modification. For the BWR examined a beneficial modification is indicated for preaction, deluge and wetpipe FPSs

in the cable spreading room. For the emergency switchgear room a beneficial modification is indicated for deluge

and wetpipe FPS. If OSAC are included all water-based

FPSs considered (both PWR and BWR) would indicate a

cost-beneficial modification.

5.1.7 Alternative 7 - Upgrade the FPS Water

Quality

A cost for this modification of $1314K was estimated. The

magnitude of risk reduction that would be achieved by

improvement of water quality is not clear. Therefore a

cost/benefit ratio was not calculated. Further study on the

effect of water-based FPSs on safety-related equipment

and the improvement of the FPS water quality and its

effect may provide the data necessary to examine this.,

issue quantitatively.

NUREG-1472 18
Table 5.2 Cost/Benefit Ratio

Alternative 3

.$Kperson-SV,($K/person-rem) averted

Preaction Deluge Wetpipe CO 2 Halon

PWR

CSR 1.9 E+02

(1.9E + 00)

EDG

ESGR

CSR < 1.OE + 02

(< 1.OE + 00)

EDG

ESGR

BWR

CSR 4.8E + 01

(4.8E-01)

EDG

ESGR

CSR < .OE + 02

(< 1.OE + 00)

EDG

ESGR

Bold type indicates OSAC included.

Table 5.3 Cost/Benefit Ratio

Alternative 4

$K/person-Sv ($K/person-rem) averted

Preaction Deluge Wetpipe CO 2 Halon

PWR,

CSR 1,E+03) 3.E + 02 1.1E+03- .1.5E+ 03 4.0E + 03

(1.1E+01 (3.0E + 00) (1.1E + 01) (1.5E + 01) (4.OE + 01)

EDG

ESGR

CSR 3.2E + 02 < 1.0E + 02 3.2E + 02 4.8E +02) 2.1E+03

(3.2E-00)' (< I.OE+00) (3.2E+00) (4.8E+ 00 (2.1E+ 01)

EDG

ESGR"

BWR

CSR 2.8E + 02 7A4E 01 1.9E + 02 2.9E + 02 5.4E + 02

(2.8E + 00) (7.4E-01) (1.9E + 00) (2.9E + 00) (5.4E + 00)

EDG

ESGR

CSR 7.9E + 01 < 1.0E + 02 5.3E + 01 9.5E + 01 2.8E + 02

(7.9E-01) (< 1.OE + 00) (5.3E-01) (9.5E-01) (2.8E + 00)

EDG

ESGR

Bold type indicates OSAC included

19 NUREG-1472
Table 5.4 Cost/Benefit Ratio

Alternative 5

$K/person-Sv ($K/person-rem) averted

Preaction Deluge Wetpipe C0 2 Halon

PWR

CSR 7.3E + 02

(7.3E+00)

EDG

ESGR 6.8E + 02

(6.8E + 00)

CSR 1.8E + 02

(I.8E + 00)

EDG

ESGR 1.E + 02

(1.5E + 00)

BWR

CSR 1.8E + 02

(1.8E + 00)

EDG

ESGR 1.8E + 02

(1.8E + 00)

CSR 4.4E + 01

(4.4E-01)

EDG

ESGR 4.0E + 01

(4.0E-01)

Bold type indicates OSAC included

5.1.8 Alternative 8 - Replace Deluge with

Preaction Sprinkler FPS

This modification applies to all three plant areas examined and would require one FPS replacement in the

*CSR and two in both the EDG and the ESGR area.

However, a cost/ benefit was not performed for the ESGR

since a preaction FPS was not part of the configurations

studied based on Appendix D of Reference 9. The cost for

this modification was estimated to be $25K for one plant

area. The reduction in risk associated with this modification is the difference in risk between a deluge FPS and a

preaction FPS. A cost-beneficial modification is indicated

for both the cable spreading room and the emergency

diesel generator area without OSAC included for both the

PWR and BWR examined. Thble 5.6 presents the cost/benefit results for this modification.

5.1.9 Alternative 9 - Replace Electrical

Cabinet with a Cabinet Designed to

Prevent Water Intrusion

This modification was examined on a plant area basis

assuming two cabinets in each of two EDG areas, two

cabinets in the CSR and ten cabinets in each of two ESGR

areas. These numbers were estimated based on plant

walkdowns and may vary for specific plants. The cost for

this modification was estimated to be $25K for one cabinet

replacement. The reduction in risk associated with this

modification is the elimination of all water-based core

damage frequency contributions except for Root Cause

12. For the deluge FPS this modification appears to be

cost-beneficial in the cable spreading room and emergency diesel generator areas for both the PWR and BWR

examined without OSAC included. If OSAC are included,

this modification also appears to be beneficial for a wetpipe FPS in an EDG area in a BWR. The results of the

cost/benefit are presented in Table 5.7.

NUREG-1472 20
Table 5.5 Cost/Benefit Ratio

Alternative 6

$K/person.Sv ($K/person-rem) averted

Preaction Deluge Wetpipe CO 2 Halon

PWR

CSR 2.ME + 02 2.ME + 02 2.ME + 02

(2.OE + 00) (2.OE + 00) (2.0 E + 00)

EDG

ESGR < 1.OE + 02 < .OE + 02

(< 1.0E + 00) (< 1.OE + 00)

CSR

EDG

ESGR

BWR

CSR 4.8E + 01 4.8E + 01 4.8E + 01

(4.8E-01) (4.8E-01) (4.8 E-01)

EDG

ESGR < 1.OE + 02 < 1.OE + 02

(< 1.OE + 00) (< 1.0 E + 00)

CSR

EDG

ESGR

Bold type indicates OSAC included

Table 5.6 Cost/Benefit Ratio

Alternative 8

$K/person.Sv ($K/person-rem) averted

Preaction Deluge Wetpipe CO 2 Halon

PWR

CSR 4.5E + 01

(4.5E-01)

EDG 3.3E +01

(3.3E-01)

ESGR

CSR < 1.OE + 02

( < 1.0E + 00)

EDG < .0E + 02

.I( < L.OE + 00)

ESGR

BWR

CSR 1.1E+01

(1.1E-01)

EDG 8.8E+00

(8.8E-02)

ESGR

CSR < 1.OE + 02

( < 1.OE + 00)

EDG < 1.OE + 02

(< L.0E + 00)

ESGR

Bold type indicates OSAC included.

21 NUREG-1472
Table 5.7,,Cost/Benefit Ratio

Alternative-9

$K/person-Sv ($K/person.rem) averted .

Preaction Deluge. Wetpipe CO 2 . Halon

PWR

CSR

EDG

ESGR

CSR

EDG

ESGR

BWR

CSR

EDG

ESGR

CSR

EDG

ESGR

8.1E+04

(8.1E+02)

9.1E1+05)

(9.1E+03

8.OE + 04

(8.OE + 02)

9.1E+ 05

(9.1E + 03)

9.1E+01

(9.1E-01)

6.7E + 01

(6.7E-01)

1.3E + 03

(13E+01)

< 1.OE + 02

(< 1.OE + 00)

< .OE + 02

(< 1.E + 00)

6.8 E+02

(6.8E + 00)

2.3E + 01

(2.3E-01)

1.8E + 01

(1.8E-01)

3.2E + 02

(3.2E + 00)

< .OE + 02

( < .OE + 00)

< 1.0E + 02

( < 1.0E + 00)

1.7E + 02

(1.7E + 00)

1.6E + 04

(1.6 E+02)

3.4E + 05

(3.4 E+ 03

1.7E + 04

(1.7 E + 02)

1.SE + 04

(1.5 E+02)

3.4E + 05

(3.4 E+03

1.6E + 04

(1.6E + 02)

1.3E + 02

(1.3E + 00)

1.OE + 05

(1.0E + 03)

4.2E + 03

(4.2E + 01)

1.3E + 02

(1.3E + 00)

1.8E+01

(1.8E-01)

4.E + 03

(4.OE + 01)

2.ME + 04

(2.OE + 02)

2.4E + 05

(2.4E + 03)

2.0E + 04

(2.0E + 02)

2.4E + 05

(2.4E + 03)

Bold type indicates OSAC included.

5.1.10 Alternative 10 - Seismically. Anchor

Safety-Related Cabinets Susceptible to

Tipping/Sliding Failure

This modification applies to the CSR and ESGR areas

and the cost appropriately reflects the number of electrical cabinets in each area. It is recognized that some of the

cabinets may already be seismically anchored, but for the

purposes of this estimate it is assumed, that all -cabinets

will require the seismic anchoring modification. The cost

for seismically anchoring one cabinet is estimated to be

$76K. The reduction in risk associated with this modification is the elimination of Root Cause 12 scenarios in the

CSR and ESGR areas, Thble 5.8 presents the cost/benefit

results for this modification. This modification appears.to

be beneficial when OSAC are included for a preaction,

deluge and wetpipe FPS in the CSR for.the BWR examined.

5.2 Frontfit Analysis

The plant modifications presented in the cost/benefit..

analysis as part of the Generic Issue 57 generic plant,.

analysis were intended to be backfits for existing plants

and as such these proposed plant modifications, were determined from'the insights provided by the three'individi'-'

al plant analyses and the generic plant analysis. However,

some of these plant modifications may be considered as

frontfits as part of the Advanced Light Water Reactor:.

(ALWR) design program. All of the modifications consid-J,

ered as frontfits would avoid cost contributions for health

physics support, radiation exposure, waste disposal and.

removal labor. Licensee and NRC costs were not recalculated and are not expected to differ significantly from the',

backfit estimates. The following subsections specifically

discuss the proposed plant modifications, as applicable,

for frontfits.

NUREG-1472 22
Table 5.8 Cost/Benefit Ratio

Alternative 10

$K/person-Sv ($K/person.rem) averted

Preaction Deluge Wetpipe CO2 Halon

CSR 7.6E + 02 7.6E + 02 7.6E + 02 1.0E + 03 2.8E + 03

(7.6E + 00) (7.6E + 00) (7.6E+ 00) (1.OE + 01) (2.8E + 01)

EDG

ESGR 7.6E + 03) 7.6E + 03 9.5E + 03 2.2E + 04

(7.6E + 01 (7.6E + 01) (9.5E + 01) (2.2E + 02)

CSR 2.2E + 02 2.2E + 02 2.2E + 02 4.7E + 02 2.3E + 03

(2.2E + 00) (2.2E + 00) (2.2E + 00) (4.7E + 00) (2.3E + 01)

EDG

ESGR 7.1E+03 7.1E+03 9.OE + 03 2.1E + 04

(7.1E + 01) (7.1E+01) (9.OE + 01) (2.1E + 02)

CSR 1.9E + 02 1.9E + 02 1.9E + 02 2.5E + 02 6.9E + 02

(1.9E + 00) (1.9E + 00) (1.9E + 00) (2.5E + 00) (6.9E + 00)

EDG

ESGR 2.1E1+03 2.1E1+03 2.5E + 03 6.1E+03

(2.IE+01) (2.1 E + 01) (2.5E + 01) (6.1E+01)

CSR 5.4E + 01 5.4E + 01 5.4E + 01 1.1E+02 5.6E + 02

(5.4E-01) (5.4E-01) (5.4 E-01) (I.IE+ 00) (5.6E + 00)

EDG

ESGR 1.9E + 03 1.9E + 03 2.4E + 03 5.9E + 03

(1.9E + 01) (1.9 E + 01) (2.4E + 01) (5.9E + 01)

Bold type indicates OSAC included.

5.2.1 Alternative 1 - No Action

Under this alternative there would be no new regulatory

requirements. Consistent with existing regulations, this

alternative does not preclude an applicant under 10 CFR

Part 50 or Part 52 from proposing to the NRC staff design

changes intended to enhance the reliability or operability

of the fire protection systems and their components on a

plant design-specific basis.

5.2.2 Alternative 2 - Upgrade a FPS

Actuation Controller with Seismically

Qualified Printed Circuit Boards

Because of concerns and industry experience with relay

chatter during seismic events it may be prudent to investigate replacing existing relays in the FPS, controller cabinets with printed circuit boards. Based on plant specific

walkdowns many types of FPS actuation relays were

found. For some plants, mercury Wetted relays for FPS

actuation and/or the annunciation of alarms and isolation

of room cooling was found. Given a seismic event, there is

a high likelihood of actuation for some of these relay

types. The cost determined for this design is for one FPS

actuation controller and would, eliminate Root Cause 8

(relay chatter in a seismic event) contributions in the area

of installation only. If this modification is performed as a

frontfit it would be included as part of the overall FPS

design. However, it is assumed that the Costs associated

with including this relaytype as part of a new design would

be similar to that of the backfit costs minus the costs of the

original relay. The total cost of this relay 'ranges from

$13,000 to $17,000.

5.2.3 Alternative 3 - Replace Smoke Detector

Actuated FPS with a Heat Detector

Actuated FPS

Designing a FPS to actuate on heat detectors rather than

smoke detectors will eliminate contributions from Root

Cause 1, 7 and 10 which are smoke detector specific. The

cost for this system as a frontfit would be similar to the

costs estimated for the components considered as part of

23 NUREG-1472
an existing plant backfit. The cost to implement this option would range from $78,000 to $105,000.

5.2.4 Alternative 4 - Reroute Safety-Related

Cables

The intent of this modification is to reroute one set of

redundant cabling to remove it from a common mode

failure vulnerability. This plant modification would not be

considered as a frontfit and would be proposed as part of

the new plant design.

5.2.5 Alternative 5 - Seismically Qualify the

CO2 Tank, Outlet Piping, and Battery

Rack

A seismic walkdown and subsequent fragility analysis for a

typical CO2 system found a high likelihood of suppressant

agent diversion given an earthquake. Failure to the tank

or its outlet piping or the FPS battery dominated the

overall probability of failure of the CO 2 system. This potential plant system design would seismically qualify the

CO 2 tank, battery, and its immediate outlet piping. The

overall costs for this frontfit would not differ from the

backfit costs significantly. The total cost for this design

ranges from $97,000 to $131,000.

5.2.6 Alternative 6 - Seismically Qualify a

FPS Battery Rack

The cost for this system as a frontfit would be similar to

the costs estimated for an existing plant backfit. The total

cost for this upgrade ranges from $35,000 to $47,000.

5.2.7 Alternative 7 - Upgrade the FPS Water

Quality

Water is the most frequently used FPS agent at nuclear

power plants. The most common delivery systems are

pre-action sprinklers, deluge wet-pipe sprinklers, drypipe sprinklers and standpipe hose systems. It has been

postulated that improvement of water quality will reduce

the potential for damage to safety-related components

from exposure to water suppressant. To address this concern, a plant design to upgrade the FPS water quality is

considered. For this design a water purification system

would be required along with the associated piping and

valves and a storage tank. Additional costs to be considered for a frontfit would be the FPS sprinkler piping and

sprinkler heads. The total cost for this system design

would range from $1,174,000 to $1,577,000.

5.2.8 Alternative 8 - Replace Deluge with

Preaction Sprinkler FPS

This modification would not be considered a frontfit, but

part of an overall FPS design. However, it is anticipated

that forthe components of the FPS analyzed as part of the

backfit analysis utilized as part of the new FPS design the

costs will be similar.

5.2.9 Alternative 9 - Replace Electrical

Cabinet with a Cabinet Designed to

Prevent Water Intrusion

One study (Ref. 11) has reported that electrical equip'-

ment failure modes related to water were mainly due' to

electrical shorting and long term corrosion. Wherever

water intrusion is possible the equipment could be expected to fail through shorting, grounding, tipping 'of

overcurrent devices, physical damage due to the velocity

of direct hose steams or long term corrosion causing potential failure of electro-mechanical parts. These failure

modes were found to be dependent on the National Electrical Manufactures Associated (NEMA) rating or the

configuration of the electrical enclosure. The appropriate

NEMA rating to preclude water intrusion can potentially

eliminate failures in electrical equipment due to water

based FPSs. Enclosures that have NEMA rating of 1 and'5

are subject to water intrusion under all water spray conditions applicable for this study. Enclosures with NEMA

ratings of 2, 3, 3R, 3S, 4, 4S, 6, 6P, 11, 12, 12K and 13 are

expected to prevent water intrusion under direct hose

stream and splashing water spray. Only 1those enclosures

with a NEMA rating of 6 and 6P are expected to prevent

water intrusion under temporary submersion due -to

flooding. The intent of this modification is to replace

existing safety-related electrical cabinets and enclosures

with NEMA spray-proof rated enclosures.

The cost for this system as a frontfit would be similar to

the costs estimated for an existing plant backfit with the

exception of the labor for the removal of old electrical

cabinets. The cost to implement this modification as, a

frontfit would range from $22,000 to $30,000 per cabinet.

5.2.10 Alternative 10 - Seismically Anchor

Safety-Related Cabinets Susceptible to'

Tipping/Sliding Failure

In a seismic event energized cabinets present a potential

source for fire. Although it is assumed that in a seismic

event offsite power will be list, thus deenergizing many

electrical cabinets, there will be a number of safetyrelated cabinets energized by alternative power sources

(batteries, diesel generators). These energized cabinets

are susceptible to tipping/ sliding failure possibly leading

to a fire. It is assumed that this modification would eliminate the potential for seismically induced fires due to the

tipping/sliding failure of an energized electrical cabinet.

The cost for this system as a frontfit would be similar to

the costs estimated for an existing plant backfit. The total

cost for this design ranges from $67,000 to $91,000.

Additional technical insights for the ALWR and the potential risks associated with the actuation of fire protection systems are given in some detail in Chapter 6 of

Reference 9.

NUREG-1472 24
6 DECISION RATIONALE

Generic Issue 57, "Effects of Fire Protection System Actuation on Safety-Related Equipment", was identified in

1982 (Refs. 1,2) as a result of a number of precursor events

showing that safety-related equipment subjected to fire

protection system (FPS) water spray could be rendered

inoperable. These precursor events also indicated numerous spurious FPS actuations initiated by operator testing

errors or by maintenance activities, e.g., welding, and

steam or high humidity in the vicinity of FPS detectors or

control circuitry and components. An NRC memorandum

(Ref.2), issued on January 28, 1982, provided additional

examples of FPS actuation interactions and suggested

that all types of FPS, e.g., water, halon, carbon dioxide

and other chemicals be included in a review of their safety

significance and possible corrective steps.

FPS actuations that result in adverse interactions with

plant systems needed to achieve safe plant shutdown or to

mitigate a postulated accident reduce the availability of

such systems. This concern is accentuated when common

cause initiators and common mode failures of safetyrelated equipment are considered. Examples of common

.cause initiators include earthquakes, smoke and heat intrusion into multiple fire zones, and fire suppressant intrusion into multiple fire zones affecting several safetyrelated systems. Examples of common mode failures of

safety-related systems and/or auxiliary systems supportin g safety-related systems include electrical shorts in instrument cabinets and electrical power distribution centers, CO2 ingress into the fresh air intake of emergency

diesel-generator sets, and cold CO 2 induced thermal

stresses and cracking of station battery casings, with loss

of offsite power during an earthquake. It should be noted

that a number of common cause initiators and common

mode failures are not mutually exclusive and they may be

part of a single event sequence.

Four plants representative of the various designs of currently operating nuclear power plants were evaluated as

part of this issue (Refs. 4-8). Furthermore, a generic evaluation of this issue (Ref. 9) was performed taking into

account the insights from the technical findings of these

four evaluations, as well as design and plant layout information of a large number of operating plants collected for

thispurpose. An extensive review of operational experience involving actuations of fire protection systems was.

performed prior to the analytical assessments of risk associated with this issue. The review of the operational experience and the aforementioned analyses showed the following:

o, 0.15 inadvertent FPS actuations/RY

0 0.02 advertent FPS actuations/RY

* 37% of all actuations damaged some equipment

* 20% of all actuations resulted in a plant transient

and reactor trip

* Core Damage Frequency (CDF) contributions from

GI-57 root causes for the four individual plants evaluated were estimated to be in the range of

7.3E-06/RY to 5.6E- 05/RY.

* Dominant risk contributors are associated with seismic/FPS and seismic/fire interactions resulting in sequences involving station blackout and small

LOCAs

Both of these categories of dominant sequences are currently being addressed by the Individual Plant Examination of External Events (IPEEE) program. Generic Letter 88-20, Supplement 4, June 27, 1991 (which initiated,

the IPEEE) states:

"The walkdown procedures should be specifically

tailored to assess the remaining issues identified in.

the Fire Risk Scoping Study: (1) seismic/fire interactions, (2) effects of fire suppressants on safety equipment. (paragraph 4.2).

The staff notes that any complete assessment of seismic/

fire interactions would necessarily include seismic effects

*on manual firefighting, since automatic fire suppressant

systems are not seismically qualified, thus increasing the

potential need for effective manual firefighting during

seismic events.

NUREG-1407 ("Procedural and Submittal Guidance for

the IPEEE for Severe Accident Vulnerabilities", June,

1991) reiterates the above (in the first paragraph of section 4) and also adds:

"The'use of an existing fire PRA for the internal fires

IPEEE is acceptable provided the PRA reflects the

current as-built and as-operated status of the plant

and the licensee addresses the deficiencies of past

PRAs that are identified in the Fire Risk Scoping

Study (NUREG/CR-5088). Deficiencies may include the use of low conditional failure probabilities

for dampers and penetrations, no consideration of

damage from the use of fire suppressants, inappropriate estimates of the effectiveness of manual fire

fighting, and no consideration of seismic/fire interactions." (paragraph 4.2).

Also in NUREG-1407, Appendix D, section 6 ("Internal

Fires") the staff response to question 6.2 states:

"The procedurally directed walk-downs associated

with internal fires vulnerability evaluation can be

planned as part of the seismic walk-downs that would

specifically look for the seismic-induced fire

25 NUREG-1472
vulnerability issues. The idea is to first identify those

areas that could be vulnerable so that they can be

brought into focus during the walkdown.

"For example, if a plant didn't have its diesel fuel

tank strapped down properly one could postulate a

large fuel source for fire as a result of a seismic

event. Other similar seismic/fire interactions were

summarized in Section 7 of NUREG/CR-5088."

In addition, for performance of the IPEEE, many'licensees will also be using the Fire-induced Vulnerability Evaluation (FIVE) methodology developed by the Electric

Power Research Institute (EPRI) as described in "Fire-induced Vulnerability Evaluation (FIVE)", EPRI TR100370, April, 1992. Following a description of the three

basic Phases of FIVE, that document states:

In addition, there is a discussion of several potentially risk significant items that were identified in the

NUREG/CR-5088, "Fire Risk Scoping Study", that

should also be considered in performing FIVE. (Section 7.0) [emphasis added].

Section 7.0 of NUREG/CR-5088 is a description of the

Sandia Fire Risk Scoping Study Evaluation which includes

discussion of the dominant risk contributors to

GSI-57-related events, including Seismic/Fire interactions (seismically induced fires, seismic actuation of fire

suppression systems, and seismic degradation of fire suppressant systems), and Manual Fire Fighting Effectiveness. The following seismic/fire interactions are given as

examples of the situations to be considered by the IPEEE:

unanchored CO 2 or Halon tanks, possible relay chatter in

fire protection system actuation systems, fire alarm systems having only a smoke-actuated alarm without a heat

or flame detector, fire pump mounts Without vibration

amplitude stops, cast iron fire mains, inadequate anchoring of electrical cabinets and inadequate slack in the wires

leading to such cabinets (to avoid sparks from tight wires),

unanchored high-pressure gas bottles, potential interactions between sprinkler system heads and adjacent pipes,

and presence of mercury switches in fire suppression and

detection systems (such switches should be replaced with

alternate, seismically insensitive switches).

For the "generic" evaluation, after subtraction of CDF

contributions from GSI-57-related events involving:

(a) Seismic/Fire: - seismic induced fire plus seismic induced suppressant diversion. The unsuppressed fire and/or the diverted suppressant

incapacitate safety related equipment needed

to mitigate effects of the seismic event; and

(b)' Seismic/FPS: - seismic induced actuation of the

FPS. Released suppressant damages safetyrelated equipment needed to mitigate effects

of the seismic event,

which are being emphasized by the IPEEE, the mean

CDF of the remaining contributors is less than

1.OE-05/RY which does not justify a generic backfit.

The risk reduction estimates, cost/benefit analyses, and

other insights gained during this effort have shown that

implementation of the recommendations contained in

this report can significantly reduce risk, and 'that these

improvements can be warranted in accordance with the

backfit rule, 10 CFR 50.109(a)(3). However, plant specific

analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for this

purpose are underway as part of the Individual Plant Examination of External Events (IPEEE) program.

Hence, we concluded that GSI-57 should be resolved with

no new regulatory requirements generically imposed for

existing plants.

For new plants, the enhanced safety requirements given

in SECY-90-016, "Evolutionary Light Water Reactor

(LWR) Certification Issues and their Relationship to Current Regulatory Requirements", January 12, 1990, allow

the same GSI-57 resolution to be applied (i.e., with no

new regulatory+ requirements imposed). This is justified

since the designers of new plants will perform plantspecific PRAs, including external events, that are equivalent to the IPEs and IPEEEs which will resolve this issue

for existing plants.

However, the reports dealing with the evaluation of this

issue, including this regulatory analysis, will be published,

or otherwise be made publicly available, so that the insights gained from this effort may be used by licensees and

applicants, and the nuclear industry in general, to take

voluntary steps for plant-specific cases, including frontfitting considerations for advanced or evolutionary reactor

designs.

NUREG-1472 26
7 IMPLEMENTATION

We concluded that GSI-57 should be resolved with no

new regulatory requirements generically imposed for existing plants, and that cost-effective modifications that

maybe desirable on specific plants will be identified by the

IPEEE program.

It is expected that the "design-specific probabilistic risk

assessment" (PRA) required for new plants by

10CFR52.47.(a).(v) will be equivalent to the IPEs and

IPEEEs being. performed for existing plants, i.e. it will

consider external events including the situations which

contribute significantly to CDF due to GI-57 related

events, and it will identify plant specific modifications that

may be. desirable.

We therefore concluded that GSI-57 should be resolved

with no new regulatory requirements generically imposed

for new plants, and that cost-effective modifications that

may be desirable on specific new designs will be identified

by the PRAs that will be performed for those designs.

8 REFERENCES

1. Memorandum from G. Lainas (NRC:NRR) to D.

Eisenhut (NRC:NRR),"Summary of the Operating

Reactor Events Meeting On January 7,1982," January 13, 1982.

2. Memorandum from C. Michelson (NRC:AEOD)'to

R. Vollmer (NRC:NRR) and E. Jordan (NRC:IE), "Effects of Fire Protection System Actuation on

Safety-Related Equipment," January 28,1982.

3. IE Information Notice 83-41, 'Actuation of Fire

Suppression System Causing Inoperability of SafetyRelated Equipment," U. S. Nuclear Regulatory

Commission, June 22, 1983.

4. J.A. Lambright, et al., "Risk Evaluation for a Westinghouse PWR, 'Effects of Fire Protection Systems

Actuation on Safety-Related Equipment' (Evaluation of Generic Issue 57)," NUREG/CR-5789,

, SAND91- 1534, December 1992.

5. J.A. Lambright, et al., "Risk Evaluation for a General Electric BWR, 'Effects of Fire Protection Systems

Actuation on Safety-Related Equipment' (Evaluation of Generic Issue 57)," NUREG/CR-5791,

SAND91- 1536, December 1992.

.6. J.A. Lambright, et al.,"Risk Evaluation for a Babcock and Wilcox PWR, 'Effects of Fire Protection

Systems Actuation on Safety-Related Equipment'

(Evaluation of Generic Issue 57)," NUREG/

* CR-5790, SAND91- 1535, September 1992..

G7. . 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.

8. Letter Report, "Seismic Risk Evaluation for a Water

Reactor, 'Effects of Fire Protection System Actuation Safety-Related Equipment'," Sandia National

Laboratories, December 1991.

9. J. A. Lambright, et al.,"Risk Evaluation of Generic

Issue 'Effects of Fire Protection System Actuation

on Safety-Related-Equipment'," NUREG/CR5580, SAND90-1507, December 1992.

10. Memorandum from Edward L. Jordan (CRGR:

. NRC) to Eric S. Beckjord (RES:NRC), "Implementation of the Safety Goals," September 1990.

11. Electric Power Research Institute, Report on the

Effects of Fire Suppressants on Electrical Components in Nuclear Power Plants, Report 01-0650-

1737, Revision C, Impell Corp., November, 1989,

12.. R. B. Worrell, "SETS Reference Manual,"

NUREG/CR-4213, S and 83-2675, May 1985.

13. W. T Wheelis, "User's Guide for a PersonalComputer-Based Nuclear Power Plant Fire Data . Base," NUREG/CR-4586, SAND86-0300, Sandia

National Laboratories, 1986.

14. Top Event Matrix Analysis Code, NUREG/

CR-4598, Sandia National Laboratories, January

1989.

15. Latin Hypercube Sampling Code, NUREG/

CR-3624, Sandia National Laboratories, March

1984.

16. Heaberlin, S., et al., "A Handbook for Value/Impact

Assessment," NUREG/CR-3568, December, 1983.

17. Claiborne, E., et al., "Generic Cost Estimates,"

NUREG/CR-4627, Revision 1, Science and Engineering Associates, Inc., February, 1989.

18. Lopez-Vitale, B., et al., FORECAST 3.0 User Manuai. SEA Report No. 89-461-04- A:1, April 1, 1990.

27 NUREG-1472
19. The Energy Economic Data Base (EEDB)

a. "Phase VIII Update (1986) Report for the Energy Economic Data Base Program," DOE/

NE-0051/1, United Engineers and Constructors, August, 1986.

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for the Energy Economic Data Base Program,"

NUREG/CR-4764, United Engineers and

Constructors, December, 1986.

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Data," Kingston, MA, R. S. Means Company, Inc.,

Annual, 1990.

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BWR Mark I Containment Improvements," NUREG/CR-5278, Science and Engineering Associates, Inc., January, 1989.

22. Memorandum from E. S. Beckjord, RES Office Letter No. 2, "Procedures for Obtaining Regulatory Impact Analysis Review and Support," November 18,

1988.

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for the Resolution of Generic Issues," May 10, 1988.

NUREG-1472 28
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Assigned by NRC, Add Vol.,

NRCM 1102, Supp., Rev., and Addendum Num3201, 3202 BIBLIOGRAPHIC DATA SHEET bers, If any.)

(See Instructions on the reverse) NUREG-1472

2. TITLE AND SUBTITLE

3. DATE REPORT PUBLISHED

Regulatory Analysis for the Resolution of Generic Issue 57: Effects of MONTH YEAR

Fire Protection System Actuation on Safety-Related Equipment October 1993

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

H. W. Woods Regulatory

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and

mailing address; If contractor, provide name and mailing address.)

Division of Safety Issues Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if: NRC, type "Same as above"; If contractor, provide NRC Division, Office or Region,

U,S, Nuclear Regulatory Commission, and mailing address.)

Same as above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Actuation of Fire Protection Systems (FPS) in Nuclear Power Plants have resulted in adverse interactions with

equipment important to safety. Precursor operational experience has shown that 37% of all FPS actuations damaged

some equipment, and 20% of all FPS actuations have resulted in a plant transient and reactor trip. On an average,

0.17 FPS actuations per reactor year have been experienced'in nuclear power plants in this country. This report

presents the regulatory analysis for GI-57, "Effects of Fire Protection System Actuation on Safety-Related Equipment". The risk reduction estimates, cost/benefit analyses, and other insights gained during this effort have shown

that implementation of the recommendations contained in this report can significantly reduce risk, and that these

improvements can be warranted in accordance with the backfit rule, 10 CFR 50.109(a)(3). However, plant specific

analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for this purpose are underway as part of the Individual Plant Examination of External Events (IPEEE) program.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers In locating the report,) 13. AVAILABILITY STATEMENT

Unlimited

g ,eneric issue 14. SECURITY CLASSIFICATION

fire protection system (This Page)

regulatory analysis Unclassified

safety-related equipment (This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89)


Federal Recycling Program

NUREG-1472 REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 57:

EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON SAFETY-RELATED EQUIPMENT

OCTOBER 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

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