Information Notice 1996-45, Potential Common-Mode Post-Accident Failure of Containment Coolers

Potential Common-Mode Post-Accident Failure of Containment Coolers

ML031060112
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	08/12/1996
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-96-045, NUDOCS 9608090031
Download: ML031060112 (10)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001 August 12, 1996 NRC INFORMATION NOTICE 96-45: POTENTIAL COMMON-MODE POST-ACCIDENT FAILURE OF

CONTAINMENT COOLERS

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees to a potential common-mode post-accident failure of

containment coolers. 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

On February 13, 1996, the licensee for Diablo Canyon determined that the

containment fan coolers were susceptible to component cooling water flashing

in the cooler unit cooling coils during a design-basis loss-of-coolant

accident (LOCA) with a concurrent loss of offsite power or with a delayed

sequencing of safety-related equipment. Diablo Canyon has five fan coolers

per unit. The containment fan coolers are designed to remove heat from the

containment atmosphere during normal operation. The fan coolers are also

automatically initiated engineered safety features, which help maintain

containment integrity by reducing post-accident containment pressure by the

condensation of steam and the removal of heat in conjunction with operation of

the containment spray system. The fan coolers are cooled by component cooling

water under both normal and accident conditions.

During a postulated design-basis LOCA with a concurrent loss of offsite power, the component cooling water pumps and the fan cooler fans would lose power.

The component cooling water flow stops almost immediately, while the fans

coast down over a period of minutes. The first fan cooler will restart on

slow speed approximately 22 seconds after the loss of offsite power. The

component cc'ling water pumps will restart 26 to 30 seconds after the

accident. In this scenario, the high temperature containment atmosphere would

be forced across the cooling coils for several seconds with no component

cooling water flow. This process may cause the stagnant component cooling

water to boil, thereby creating steam voids in the cooling coils and the

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IN 96-45 August 12, 1996 component cooling water piping. When the component cooling water pumps

restart, the pumped liquid would collapse the steam voids and may create

significant hydrodynamic loads (waterhammer). The hydrodynamic loads may

adversely affect the integrity of the fan coolers or the associated component

cooling water piping. This problem may cause the fan cooler pipes to rupture

and could cause a significant loss of component cooling water inventory, which

could threaten cooling to emergency core cooling system components. The

licensee for Diablo Canyon evaluated this potential failure mechanism and has

installed a nitrogen pressurization system on the component cooling water head

tank to increase the margin to boiling.

On July 22, 1996, the licensee for Haddam Neck declared all four of its

containment air recirculation fans inoperable and initiated a plant shutdown

The air recirculation fans are the only credited heat removal system for post- accident containment heat loads, and the associated cooling coils are cooled

by the station service water system. The fans were declared inoperable after

the licensee review of the consequences of two-phase service water flow

following a LOCA concurrent with a loss of normal power. The licensee

analysis predicted hydrodynamic loads that exceeded the service water piping

and support structural limits.

Discussion

At many plants, containment fan coolers provide a significant safety function

in reducing containment pressure and removing heat from the containment

building. The postulated failure scenario could cause a common failure of all

the containment coolers, thereby potentially challenging the integrity of the

containment building. In certain cases, failure of the cooling water system

piping could cause a potential containment bypass flow path and may jeopardize

cooling to other safety-related loads.

An individual plant vulnerability to these postulated failures is dependent on

a number of factors. The coastdown rates of the cooling water flow and the

cooler fans, the operating pressure and pressure decay rate of the cooling

water system, the timing of cooling water pump restart, the containment

temperature profile during the design-basis accident, and other site-specific

parameters have an effect on facility vulnerability to this potential failure

mode. This failure mode would also be applicable to the design-basis main

steam line break in containment.

Westinghouse issued Nuclear Safety Advisory Letter NSAL-96-003, "Containment

Fan Cooler Operation During a Design Basis Accident," to alert their customers

to this concern. NSAL-96-003 is provided as an attachment to this information

notice.

I--1 IN 96-45 August 12, 1996 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: James Tatum, NRR John Tappert, NRR

(301) 415-2805 (301) 415-1167 Internet: jetl@nrc.gov Internet: jrt@nrc.gov

Alan Levin, NRR Howard Wong, Region IV

(301) 415-2890 (510) 975-0296 Internet: ael@nrc.gov Internet: hjw@nrc.gov

William Raymond, Region I

(203) 267-2571 Internet: wjr~nrc.gov

Attachments:

1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96

2. List of Recently Issued NRC Information Notices

4-11,gC-411 'LJ 41-1 k-0 t

Westinghouse

Energy NUCLEAR SAFETY ADVISORY LETTER

Systs

-a Business

Usk

TIS [S A NOTIFICATION OF A RECENTLY IDENTIFIED POTENTIAL SAFEY ISSUE PERTANrNG TO BASIC

COMPONENU SUPPLIED BY WESTINGHOUSE THIS INFORMATION 1S BEING PROVIDED TO YOU SO THAT A

REVIEW OF THIS ISSUE CAN BE CONDUCTED BY YOU TO DETERMINE IF ANY ACTION IS REQUIRED.

P.O. Box 335. Pkubuqgb. PA 152304355 Subject: Containet Fan Cooler Operation During a .Dosipn Basis Accidl?{ umber. NSAL-96-003 Basic Coponent: Coota-eat Fan Coole Date: 6/20/96 Plaits: Westinghouse and MHI NSSS Plants

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. Transfer of ormation Pursuant to 10 CFR 21.21(b) Yes O

- 'Advisory Eformaion Pursuant to 10 CFR 21.21(cX2) Yes M

.Refcreace:

• -UMM.AR

The purposc of this loe is to provide Westintouse NSSS Ownens information regarding the potential

paformace of some containment fan cooler units (CFCUs) during accidenit conditions which may

compromise the containment cooling system capability.

This Westinghouse Nuclear Safety Advisory Letter deals with the potential susceptibility for the cooling

water of cain CFCUs to flash to steam during a desin basis Lou of Coolant Accident (LOCA) with

ether a cocurrent LOSS of Offsite Power (LOOP) or a delayed sequencing of safety-related equipment.

The applicability of this issue to a given nuclear unit depends on plant specific aspects of the containment

cooling sysem desip and its heat removal system. The Nuclear Regulatory Commission has been

Rotified by _ LER (Reference 1).

Additional iafornation, if required, may be obtained from the originator. Telephone 412-374-5750.

Originator(s):

J. T. Crane W. R. Ricc. Interim Manager

Regulatoiy & Licensing Initiatives Regulatory & Licensing Initiatives

AlTACHMM 1 Ul *sWrW Pff Sheet I of 3

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ISSUE DESCRIPTION

The potenial for this condition wu discovered during an %vestivgeto into Component Cooling Water

(CCW) system design issues by a Westinghouse NSSS Owners The CCW System provides cooling

water directly to the CFCUs in that plant's design. Depending on the specific design conditions, it

may be applicable to plants that utilize cooling waler systems other thin the CCW for their CFCUs

This issue concerns a type of containment cooling deaig that has a single set of CFCUs for heat

removal during both nonmal operating conditions and accident conditis. In the Licensee's design the

CFCUs arc equipped with a swo-speed blower that operates at about 1200 rpm for normal operation

and about 600 rpm for accident conditions. Cooling water to the heat exchanger coils for both normal

and accident operation is supplied by the CCW sytem. The License's containmcn analysis ausumas

the fan coolers operate during postulated accidents, removing beat so as to maintain containment

pressure within design bais limits.

During a postulated LOOP coincident with a design basis LOCA, power is lost to bpth the CCW

pumps and the CFCU blowan. The CCW pumps aem calculated to coast down to 0 rpm within 1-2 seconds while the CFCUI blowers ar calculated to coast down firm a aomiasl speed of 1200 rpm to

600 rpm in approximately 30 seconds and then to 0 rpm an estimated 700 second laser. Accountiig

for diesel-generator start time'seand emergency buss loading sequence, the CFCU blowers are rc-

• energized approximiaeij 20 seconds afler the initiation of the LOOP, and the CCW pumps at re- energized about 30 seconds after the initiation of the LOOP. This timing of events provides for hot, steam-laden cont ent ai to be drawn over the heat exchanger coils at a elatively high velocity for

30 seconds before cooling liquid flow is reestablished to the coils.

The high heat content of containment atmosph=r under accident conditions being drawn over the

CFCU heat exchanger coil with no pumped liquid flow kw been calculated to resit in steaming of the

stationary liquid inventory of the coils. When the CCW pups are re-nergizod, the pumped liquid

flow acts to collapse the steam void in Oe CFCU piping. Collapsing &a steam void is predicted to

mrIlt iG.a wzasAsmer with nfficient energy and force that may impact the integrity of either the

beat exchanger coil or its associated piping, resulting in a lost of inteWy of the CCWs rystm.

To prevent flashing of the CFCU cooling woter, the Licensee has instafled a nitrogen pressurization

system on the CCW system.

TECHNICAL EVALUATION

The following discussion provides several items to consider in evaluating your CFCUs.

In general a plant u has CFCUs with a 2 speed blower motor and a diesel loading sequence that

results in water flow through the cooler coil later than 15 seconds after a design basis large break

LOCA, may have a similar situation. However, there arn other plant specific aspects that can impact

CFCU functioning. The containment cooling design in this issue has a singk set of fan coolens that

are used for containment cooling during both normal and accident conditions. Flashing and

subsequent void collapse may not apply if a plant has separate sets of equipment where the accident

CFCUs arc not operating prior to a design basis accident. These accident related fan coolers would be

at containment ambient conditions with no forced convective et transfer at the beginning of the

13i7AMWFWIN

Sheet 2 of 3

I

accident. Since their fmns would not be operating u the beginning of the Postulated accident (as

compated to a single set f fa cooles which would be coasting down) the heat release from a design

basis accacent wouia be ttisfurred to the coil cooling water ju a much slower rate.

Flashing of the CFCU cooling water may not occur in a specific plant design because of factors such

as hea transfer capability, saturation pressure of the cooling systcm. coutdown time of the fan

blower, and coutdown time of the cooling water pump(s). Also, the timing of when ECCS eQuipment

will start in a design basis accident can determine if the CFCUs do or do not have stagnant cooling

water.

In summary, the issue is wbether safety related CFCUs will function as intended during a design basis

event, if they am postulated to experience flshing of the water in the cooling coils. If Rubing can be

postulated, then the issue of whether the CFCUs will maiuntan euctrl integrity during subsequent

void collapse should be addrssod.

ASSESSMENT OF SAFETY SIGNMCANCE .

Westinghouse canot deteaine the potential for flashing in the CFCUr on a generic basis because of

the various plant specific designs for the containment cooling system and situation specific heat

umnsfer conditions. Thus &* mott effective action that can be taken is a notification to all

Westinghouse plants.

REEORTABILrY

The NRC is awarm of this imue via an LER (Reference 1).

RECOMENDEDACTINS

Licensees should review d* containment cooling system to determine if their safety related

contaL-=e-t fan ccolerv am nsceptible to cooling water void formation and subsequent void collapse

and waterhammer during a design bais accident.

REFERENCE

1. Pacific Gas A Etcic letter. PO&E DCL-96-097, April 26, 1996 to U. S. Nuclear Regulatory

Commission, 'Licec Evet Report 1.96-005-00 (Voluntary) Potential for Flushing in

Containment Fin Cooer Units

Sheet 3 of 3

I

Attachment

IN 96-45 August 12, 1996 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

%_

s41TOr.ASDPAP

A c -crssq

.UeiUFU WTA_ Urn

n aa, raMLAr.__^__ Z/U?/vb All holders of Ls or CPs

irip breaker trom Lracking for nuclear power reactors

of Phenolic Material in

secondary contact assembly

96-43 Failures of General 08/02/96 All holders of OLs or CPs

Electric Magne-Blast for nuclear power reactors

Circuit Breakers

96-42 Unexpected Opening of 08/05/96 All holders of OLs or CPs

Multiple Safety Relief for nuclear power reactors

Valves

96-41 Effects of a Decrease in 07/26/96 All holders of OLs or CPs

Feedwater Temperature on for pressurized water

Nuclear Instrumentation reactors

96-40 Deficiencies in Material 07/25/96 All holders of OLs or CPs

Dedication and Procurement for nuclear power reactors

Practices and in Audits of

Vendors

96-09, Damage in Foreign Steam 07/10/96 All holders of OLs or CPs

Supp. 1 Generator Internals for pressurized-water

reactors

96-39 Estimates of Decay Heat 07/05/96 All holders of OLs or CPs

Using ANS 5.1 Decay Heat for nuclear power reactors

Standard May Vary Signi- ficantly

96-38 Results of Steam Generator 06/21/96 All holders of OLs or CPs

Tube Examinations for pressurized water

reactors

96-37 Inaccurate Reactor Water 06/18/96 All pressurized water

Level Indikation and Inad- reactor facilities holding

vertent Draindown During an operating license or a

Shutdown construction permit

96-36 Degradation of Cooling 06/12/96 All holders of OLs or CPs

Water Systems Due to Icing for nuclear power reactors

OL - Operating License

CP - Construction Permit

IN 96-45 August 12, 1996 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

/s/ Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: James Tatum, NRR John Tappert, NRR

(301) 415-2805 (301) 415-1167 Internet: jetlnrc.gov Internet: jrt~nrc.gov

Alan Levin, NRR Howard Wong, Region IV

(301) 415-2890 (510) 975-0296 Internet: ael nrc.gov Internet: hjw~nrc.gov

William Raymond, Region I

(203) 267-2571 Internet: wjr~nrc.gov

Attachments:

1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96

2. List of Recently Issued NRC Information Notices

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DATE 08/08/96 08/08/96 08/08/96 Official Record Copy

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IN 96-XX

August XX, 1996 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: James Tatum, NRR John Tappert, NRR

(301) 415-2805 (301) 415-1167 Internet: jetl nrc.gov Internet: jrt@nrc.gov

Alan Levin, NRR Howard Wong, Region IV

(301) 415-2890 (510) 975-0296 Internet: ael nrc.gov Internet: hjw~nrc.gov

William Raymond, Region I

(203) 267-2571 Internet: wjr~nrc.gov

Attachments:

1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96

2. List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\JRT\CONTCLR.IN

OFFICE Contacts C/SPLB:DSSA C/PECB:DRPM D/DRPM

Km ALevin LMarsh-'i AChaffep BGrimes

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IN 96-XX

August XX, 1996 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: James Tatum, NRR John Tappert, NRR

(301) 415-2805 (301) 415-1167 Internet: Jetlnrc.gov Internet: Jrt@nrc.gov

Alan Levin, NRR Howard Wong, Region IV

(301) 415-2890 (510) 975-0296 Internet: ael~nrc.gov Internet: hjw~nrc.gov

Attachment: List OT' Recently Issued NRC Information Notices

DOCUMENT NAME E: :\JRT\4 NCR.IN

OFFIM Contacts C/SPLB:DSSA C/PECB:DRPM D/DRPM

K"E ALevin 0 + LMars AChaffee BGrimes

HWong p r>

JTatum

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DATE I / /96 4/196 / /96 / /96 Official Record Copy