Information Notice 1999-17, Problems Associated with Post-Fire Safe-Shutdown Circuit Analyses

Problems Associated with Post-Fire Safe-Shutdown Circuit Analyses
	ML031040418
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000268, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	06/03/1999
From:	Matthews D; Division of Regulatory Improvement Programs
To:	;
References
	IN-99-017, NUDOCS 9905260031
	Download: ML031040418 (14)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 3, 1999 NRC INFORMATION NOTICE 99-17: PROBLEMS ASSOCIATED WITH POST-FIRE SAFE-

SHUTDOWN CIRCUIT ANALYSES

Addressees

All holders of operating licenses for nuclear power reactors, except those who have

permanently ceased operations and have certified that fuel has been permanently removed

from the reactor.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to alert

addressees to potential problems associated with post-fire safe-shutdown circuit analysis.

These potential problems could result in a vulnerability to fire-induced circuit failures that could

prevent the operation or lead to malfunction of equipment necessary to achieve and maintain

post-fire safe shutdown. It is expected that recipients will review the information for applicability

to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no

specific action or written response is required.

Description of Circumstances

The Office of Nuclear Reactor Regulation (NRR), the NRC regional offices, and licensees have

found plant-specific problems related to potential fire-induced electrical circuit failures that could

prevent operation or cause malfunction of equipment needed to achieve and maintain post-fire

safe shutdown. Examples of problems reported by the licensees follow.

Associated Circuits

In Licensee Event Report (LER) 50-382/97-020-01 of March 16, 1998, Entergy Operations, Inc.,

the licensee for Waterford Steam Electric Station, Unit 3, reported that a fire in the switchgear

room could potentially result in the momentary loss of both trains of safety-related static

uninterruptible power supplies (UPS). The licensee attributed this condition to a combination of

the inherent design of the UPS units and unprotected UPS associated circuits that were not

separated in accordance with the regulatory requirements of Appendix R to 10 CFR Part 50.

The licensee concluded that this condition would not necessarily have prevented safe shutdown

of the plant in the event of a fire, but couEd have resulted in recurring instances of the load

breakers isolating the faults and the UPS units momentarily shutting down to clear the faults, VD(L 174g A)otlCQqO l019 O

(60031pjakd onFg4 dll

IN 99-17 June 3, 1999 which could have continued after control was transferred to the remote shutdown panel. The

licensee's corrective measures included replacing a UPS, rerouting cables, and enhancing

operator response procedures.

Cable Routing and Separation and Wirina Errors

In LER 50-266/99-002 of April 14, 1999, Wisconsin Electric, the licensee for Point Beach

Nuclear Plant, reported that it had discovered that a cable necessary to provide a plant

parameter to support post-fire safe shutdown was not routed independent of the appropriate fire

zone. According to the licensee, this condition occurred because a previous modification failed

to provide for the cable routing necessary to meet the safe shutdown evaluation. The licensee

is considering either rerouting the cable or protecting the cable with a 3-hour fire-rated barrier.

In LER 50-454/97-023-01 of March 4, 1998, Commonwealth Edison Company, the licensee for

Byron Nuclear Power Station, Unit 1, reported that, because of a wiring error that occurred

during the installation of a design change, the 1B emergency diesel generator (EDG) may not

have continued to operate in the event of a fire in certain fire zones. Specifically, a normally

connected wire was mistakenly removed and the intended wire, which remained, cross-tied

both EDG DC control power supplies. Therefore, certain fires could have resulted in the loss of

both control power circuits for the EDG, leaving the EDG without any control power and unable

to respond as designed. The licensee revised its procedures for verifying design changes, rewired the affected diesel controls, performed a wiring verification, and walked down all EDG

local panels.

Fire-induced Hot Shorts

In LER 50-341198-003-01, of July 31, 1998, Detroit Edison, the licensee for Enrico Fermi 2, reported that during an investigation of a previous LER, i discovered that a combination of

multiple fire-induced hot shorts could open a drain path and cause a loss of condensate storage

tank (CST) inventory required for dedicated shutdown. During a later review of its corrective

actions for this previous condition, the licensee determined that the corrective actions were not

adequately incorporated into the dedicated shutdown procedure. Specifically, under the revised

procedure, the valve that was susceptible to spurious opening would be closed before electrical

power was disconnected. Therefore, a fire-induced hot short could cause the valve to reopen

until it was de-energized. According to the licensee, it also previously failed to identify the

normal supply to the high pressure coolant Injection (HPCI) system and the reactor core

isolation cooling (RCIC) system as potential CST drain paths because it assumed that when the

HPCI pump was shutdown, flow through the pump would stop. Finally, the licensee previously

failed to recognize that a RCIC gravity drain path could be opened to the suppression pool by a

hot short. The licensees corrective actions included adding new manual actions to the

abnormal operating procedures, designating access routes and installing emergency lighting, reviewing other valves for susceptibility to fire-induced spurious operation, and other

Appendix R reassessment activities.

IN 99-17 June 3, 1999 Evaluations of Spurious Operations

In LER 50-255/97-008 of October 10, 1997, Consumers Power Company (CPC), the licensee

for Palisades Nuclear Power Plant, reported that it had failed to properly evaluate the potential

for spurious opening of certain service water (SW) cross-tie valves as a result of fire-induced

hot shorts. According to the licensee, had a spurious opening of the SW cross-tie valves

occurred due to the effects of a fire, loss of component cooling water inventory to the lower

pressure SW system could occur in as little as 25 seconds without any approved Appendix R

coping scenarios in place to mitigate the condition. The licensee implemented a number of

corrective actions including isolating the air supply to the SW cross-tie valves to remove the

possibility of spurious operations of the valves.

As another example, in LER 50-255/97-10 of October 30, 1997, CPC reported that its

Appendix R analysis was in error in that it accounted for the fire-induced spurious opening of

only two of the four main atmospheric steam dump valves (ASDVs) in the event of a control

room or cable spreading room fire. According to the licensee, it should have accounted for all

four main ASDVs and the turbine bypass valve spuriously opening as a result of fire-induced

hot shorts. The cause of this condition was a cable that was not identified as a field run on the

circuit diagram. According to the licensee, fire scenarios involving this condition could have

lead to fuel clad damage. As a result of a new evaluation, the licensee revised its procedures

to cope with all four ASDVs opening in the event of a fire.

Motor Operated Valve Evaluations

In LER 50-293/97-029 of January 19, 1998, Boston Edison Company, the licensee for Pilgrim

Nuclear Power Station, reported that it had determined that the shutdown cooling (SDC) suction

isolation valves were vulnerable to mechanical damage from a potential failure mode involving

hot shorts. According to the licensee, this condition could have caused spurious operation of

these SDC motor operated valves (MOVs) during a control room or cable spreading room fire, resulting in damage to either valve such that the operators could not change the position of the

valves. As a result, the SDC mode of the residual heat removal system would not be available

to support the safe cooldown of the plant. As a temporary corrective action, the licensee

opened breakers to de-energize the valves, placing them in a fail-safe, isolated condition, and

maintaining containment integrity. As a permanent corrective action, the licensee is considering

converting the temporary modification to a permanent plant modification, or modifying the

control circuits to eliminate the postulated failure mode.

Transfer and Isolation Capability

In LER 50-461/97-021 of August 25, 1997, Illinois Power, the licensee for Clinton Power

Station, reported that a control circuit fed from the 125 volt direct current (DC) control power

fuse for the Division 1 EDG feed breaker was routed through the main control room (MCR) but

did not include isolation contacts at the remote shutdown panel (RSP) transfer switch to allow

isolation of the MCR wiring/controls while operating from the RSP. According to the licensee, in

the event of a MCR fire, concurrent with the loss of offsite power, the fire could damage the

circuit, causing multiple DC ground faults, the loss of control power for the Division 1 EDG feed

IN 99-17 June 3, 1999 breaker, and a loss of power to Division 1 equipment. To resolve this condition, the licensee

rewired the control circuit at the RSP so that it included isolation contacts from the transfer

switch.

As another example, in LER 50-461/98-013 of April 20, 1998, Illinois Power reported that fire- induced electrical short circuits during a MCR fire could prevent the operation of the RCIC

system from the remote shutdown panel. According to the licensee, two cables interfacing with

the control circuit for the RCIC turbine would not isolate from the control room when the remote

transfer switch was operated. As a corrective action, the licensee will modify the interface of

two General Electric (GE) transient analysis recording system cables with the remote shutdown

panel. In a letter dated July 9, 1998, GE Informed the staff that it had reviewed the Clinton LER

and conducted a sample review of GE design documents and was not able to confirm or refute

the existence of similar problems at other plants.

Fuse/Breaker Coordination

As reported in LER 50-353/99-001, on January 19, 1999, during an engineering review at the

Limerick Generating Station, Unit 2, PECO Nuclear, the licensee, determined that a fire in the

reactor enclosure cooling water (RECW) equipment area could cause fire-induced damage to

an auto-start pressure switch in the control circuit for a RECW pump. According to the

licensee, the damage could create a hot short that would cause the pump to auto-start if the

pump control switch was in the "run" or 6auto position. In addition, the licensee determined that

the same fire could induce a fault in the 480 VAC power cable to the pump motor which could

open the load center beaker to its associated motor control center (MCC). The licensee

concluded that the fire could result in the loss of equipment required for post-fire safe

shutdown. The licensee attributed this condition to inadequate circuit breaker coordination. In

response to this condition, the licensee reset the MCC breaker settings and revised the

appropriate design calculations. The licensee will also review and upgrade, as necessary, all

safety-related breaker coordination calculations.

High ImDedance Faults

In LER 50-397/9-006-01 of July 23, 1998, Washington Public Power Supply System, the

licensee for Washington Nuclear Plant, Unit 2, reported that it had found discrepancies in low

voltage bus calculations during a review of Appendix R calculations for high Impedance faults.

According to the licensee, the calculation errors could have resulted in overloading of certain

buses due to fire-induced faults and the loss of safe shutdown capability. The licensee

conducted a review of its documentation and calculations, and revised its safe shutdown

procedures to prevent low voltage buses from becoming overloaded due to fire induced faults.

High-Pressure/Low-Pressure Interfaces

On February 27, 1997, Commonwealth Edison Company reported that during a review of the

Quad Cities Appendix R Conformance Safe-Shutdown Analysis, plant personnel discovered

that the reactor water cleanup (RWCU) system had been identified as a high-pressure/low- pressure interface requiring isolation during certain design-basis fires (LER 50-265/97-006).

The licensee stated that a postulated design-basis fire could have caused multiple spurious

, _/1 - /

IN 99-17 June 3, 1999 operations of certain RWCU system valves, potentially allowing a loss of reactor coolant

inventory in excess of design basis limits. Specifically, failure to properly isolate the RWCU

system could have prevented station operators from attaining safe-shutdown before the reactor

water level reached the top of the active fuel. The resolution of the potential RWCU blowdown

path was to ensure that the system would be manually isolated during design basis fires. The

licensee revised the appropriate safe shutdown-related documents and procedures.

Discussion

The regulatory requirements, the regulatory guidance, and the NRC staffs positions on post- fire safe shutdown are contained in various NRC documents; including General Design

Criterion (GDC) 3, "Fire protections of Appendix A to Part 50 of Title 10 of the Code of Federal

Regulations (10 CFR Part 50); 10 CFR 50.48, OFire protection"; 10 CFR Part 50, Appendix R,

"Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979";

Branch Technical Position APCSB 9.5-1, Appendix A, "Guidelines for Fire Protection for

Nuclear Power Plants Docketed Prior to July 1, 1976"; and NUREG-0800, *Standard Review

Plan." The extent to which these requirements or guidelines are applicable to a specific nuclear

power plant depends on plant age, commitments established by the licensee in developing the

fire protection plan, and the license conditions pertaining to fire protection. One of the

objectives of these requirements and guidelines is to provide reasonable assurance that fire- induced circuit failures (e.g., hot shorts, open circuits, and shorts to ground) that could

adversely affect the ability to achieve and maintain post-fire safe shutdown will not occur.

The risk and safety significance of circuit analysis problems is dependent on a number of

factors including, for example, the importance of the circuit to the post-fire safe shutdown

capability, the type and configuration of the circuit, and the potential circuit failure modes. The

NRC staff did not conduct detailed risk assessments of the events and inspection findings

discussed in this information notice. However, in view of these reports of circuit analysis

problems, and a number of similar reports, the NRC staff is treating this issue generically. The

staff is interacting with the reactor industry and other interested stakeholders to develop

effective, realistic, deterministic and risk-informed solutions to the circuit analysis issues. On

July 23, 1998, the staff conducted a public workshop to discuss with the public and the nuclear

reactor industry a variety of safety, technical, and regulatory matters associated with post-fire

safe shutdown circuit analyses. The underlying objectives of the workshop were to bound any

issues, to achieve an understanding of the industry positions on the issues, and to impart the

staffs understanding of the issues. The discussions focused on safety, technical and

regulatory issues; the assumptions that go into circuit analyses; and the terminology used to

discuss circuit analyses. To help resolve issues associated with post-fire safe shutdown circuit

analyses, the Boiling Water Reactor Owners' Group (BWROG) established an Appendix R

Committee and the Nuclear Energy Institute (NEI) formed a Circuit Analysis Issue Task Force.

NEI is developing a risk-based methodology for addressing fire-induced circuit failures. This

effort involves identifying potential circuit failure modes for specific conditions and

arrangements, developing risk-informed methods for assessing the likelihood of fires in plant- specific locations and their potential to cause multiple circuit failures, and developing

deterministic circuit failure analysis methods that could be applied to plant-specific

configurations that are not screened by the assessment of circuit failure modes or the risk

assessment. The BWROG is developing generic definitions, assumptions, and positions

related to a deterministic fire-induced circuit failure analysis methodology.

I

IN 99-17 June 3, 1999 This information notice establishes no new NRC requirements; therefore, no specific action or

written response is required by this notice. However, recipients are reminded that they are

required by 10 CFR 50.65 to take industry-wide operating experience (including information

presented in NRC information notices) into consideration when setting goals and performing

periodic evaluations. If you have any questions about the information in this notice, please

contact one of the technical contacts listed below or the appropriate NRR project manager.

David B. Matthews, Director

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Patrick M. Madden, NRR Leon E. Whitney, NRR

301-415-2854 301-415-3081 E-mail: pmm@nrc.gov E-mail: lewl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

'- 7 Attachment 1 IN 99-17 June 3, 1999 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

99-16 Federal Bureau of Investigation's 5128/99 All U.S. NRC fuel cycle, power

Nuclear Site Security Program reactor, and non-power reactor

licenses

99-15 Misapplication of 10 CFR Part 71 5/27/99 All holders of operating licenses or

Transportation Shipping Cask construction permits for nuclear

Licensing Basis to 10 CFR Part 50 power reactors

Design Basis

99-14 Unanticipated Reactor Water 5/5/99 All holders of licenses for nuclear

Draindown at Quad Cities Unit 2, power, test, and research reactors

Arkansas Nuclear One Unit 2 and Fitzpatrick

99-13 Insights from NRR Inspections 4/29/99 All holders of operating licenses

of Low-and Medium-Voltage for nuclear power reactors

Circuit Breaker Maintenance

Programs

99-12 Year 2000 Computer Systems 4/28199 All holders of operating licenses

Readiness Audits or construction permits for nuclear

power plants

99-11 Incidents Involving the Use of 4/23/99 All medical use licensees

Radioactive Iodine-131

97-15, Sup I Reporting of Errors and 4/16/99 All holders of operating licenses

Changes in Large-Break/Small- for nuclear power reactors, except

Break Loss-of-Coolant Evaluation those who have permanently

Models of Fuel Vendors and cease operations and have

Compliance with 10 CFR 50.46(a)(3) certified that fuel has been

permanently removed from the

reactor

OL = Operating License

CP = Construction Permit

IN 99-17 June 3, 1999 This information notice establishes no new NRC requirements; therefore, no specific action or

written response is required by this notice. However, recipients are reminded that they are

required by 10 CFR 50.65 to take industry-wide operating experience (including information

presented in NRC information notices) into consideration when setting goals and performing

periodic evaluations. If you have any questions about the information in this notice, please

contact one of the technical contacts listed below or the appropriate NRR project manager.

Original signed by

David B. Matthews, Director

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Patrick M. Madden, NRR Leon E. Whitney, NRR

301-415-2854 301-415-3081 E-mail: pmm@nrc.gov E-mail: lewl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\WFBkIN_HOTSH

See previous concurrence

lFFICE IPECB* l SPUB* blPLB* BG:SPLB* u :SSA*

NAME WFBurton PIMadden KSWestF JNHannon GHolahan

ATE 4120199 2/9/99 4/19/99 4/19/99 4/20/99 OFFICE BC:REXB* l D:DRIP* l D:DE* l D:NRR*

NAME LBMarsh DMatthews JRStrosnider SCollins

QATF4f7319Q Si/25/99 1)l L__n__.__ROMP

OFFICIAL RECORD COPY

IN 99-xx

May xx, 1999 This information notice establishes no new NRC requirements; therefore, no specific action or

written response is required by this notice. However, recipients are reminded that they are

required by 10 CFR 50.65 to take industry-wide operating experience (including information

presented in NRC information notices) into consideration when setting goals and performing

periodic evaluations. If you have any questions about the information in this notice, please

contact one of the technical contacts listed below or the appropriate NRR project manager.

David B. Matthews, Director

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Patrick M. Madden, NRR Leon E. Whitney, NRR

301-415-2854 301-415-3081 E-mail: pmm~nrc.gov E-mail: lewl @nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\WFB\IN_HOTSH

See previous concurrence

OFFICE l PFZECB -- -SPEB* L SPLB* BG:,SPLB' D:SSA*L

NAME WFBurton PMMadden KSWest JNHannon GHolahan

DAE 4120/X99 2YQc4Q 4119/99 _, 4#r g 4120/99q

FFICE BC:REXB Dl D:EE 5I _ I

NAME LBMarsh* *

__________tt__

_ _ _ _ 4,91/qQ 1M______________

OFFICIAL RECORD COPY

IN 99-xx

May xx, 1999 This information notice establishes no new NRC requirements; therefore, no specific action or

written response is required by this notice. However, recipients are reminded tat they are

required by 10 CFR 50.65 to take industry-wide operating experience (including information

presented in NRC information notices) into consideration when setting goals and performing

periodic evaluations. If you have any questions about the information in this notice, please

contact one of the technical contacts listed below or the appropriate NRR project manager.

David B. Matthews, Director

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical contacts: Patrick M. Madden, NRR Leon E. Whitney, NRR

301-415-2854 301-415-3081 E-mail: pmm@nrc.gov E-mail: lewl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:%WFB\INHOTSH

See nrevious concurrence

_OFFICE PECL SPLB* IBCSPLB* I D SA

NAME WFBurton IPMMadden lKSWest JNHannon GHolahan

DATE 4I2mVs 2/9/99 14/19/99 4119/99 4/20/99 F BC:REXB l D:DRIP l:DE D:NRRI Il

NAME LBMarsh DMatthews JRStrosnider SCollins

DAT I F 4/rQI _ tooI_ _ _

OFFICIAL RECORD COPY

IN 99-xx

May xx, 1999 issues. In this light, on July 23, 1998, the staff conducted a public workshop to discuss with the

public and the nuclear reactor industry a variety of safety, technical, and regulatory matters

associated with post-fire safe shutdown circuit analyses. The underlying objectives of the

workshop were to bound any issues, to achieve an understanding of the industry positions on

the issues, and to impart the staff's understanding of the issues. The discussions focused on

safety, technical and regulatory issues; the assumptions that go into circuit analyses; and the

terminology used to discuss circuit analyses. To help resolve issues associated with post-fire

safe shutdown circuit analysis, the BWR Owners' Group (BWROG) established an Appendix R

Committee and the Nuclear Energy Institute (NEI) formed a Circuit Analysis Issue Task Force.

The BWROG is developing generic definitions, assumptions, and positions related to a

deterministic fire-induced circuit failure analysis methodology. NEI is developing a risk-based

methodology for addressing fire-induced circuit failures. This effort involves identifying potential

circuit failure modes for specific conditions and arrangements, developing risk-informed

methods for assessing the likelihood of fires in plant-specific locations and their potential to

cause multiple circuit failures, and developing deterministic circuit failure analysis methods that

could be applied to plant-specific configurations that are not screened by the assessment of

circuit failure modes or the risk assessment. The BWROG is developing generic definitions, assumptions, and positions related to a deterministic fire-induced circuit failure analysis

methodology.

This information notice establishes no new NRC requirements; therefore, no specific action or

written response is required by this notice. However, recipients are reminded that they are

required by 10 CFR 50.65 tp take industry-wide operating experience (including information

presented in NRC information notices) into consideration when setting goals and performing

periodic evaluations. If you have any questions about the information in this notice, please

contact one of the technical contacts listed below or the appropriate NRR project manager.

David B. Matthews, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Patrick M. Madden, NRR Leon E. Whitney, NRR

301-415-2854 301-415-3081 E-mail: pmm@nrc.gov E-mail: lewl @ nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\WFB\INHOTSH 'Di ePioS ccfw

It

't I_:D

AME WFBurton PMMadden KSWest JN an GHolaha

DATE 14 / I Lo 9 /I /9999 / 9 4q99 OFFICE BC:REXB D:DRIP i D RRDE 1111RR111 NAME LBMarsh DMatthews JRStrosnider SCollins

QATF Ad i IQ I /QCQ 1 /C)

M A

OFFICIAL RECORD COPY

<\ 6y

-' AIN

' 99-xx

May xx, 1999 issues. In this light, on July 23, 1998, the staff conducted a public workshop to discuss with the

public and the nuclear reactor industry a variety of safety, technical, and regulatory matters

associated with post-fire safe shutdown circuit analyses. The underlying objectives of the

workshop were to bound any issues, to achieve an understanding of the industry positions on

the issues, and to impart the staffs understanding of the issues. The discussions focused on