Information Notice 1998-31, Fire Protection System Design Deficiencies and Common-Mode Flooding of Emergency Core Cooling System Rooms at Washington Nuclear Project Unit 2

Fire Protection System Design Deficiencies and Common-Mode Flooding of Emergency Core Cooling System Rooms at Washington Nuclear Project Unit 2

ML031050080
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, 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, Crane
Issue date:	08/18/1998
From:	Roe J Office of Nuclear Reactor Regulation
To:
References
IN-98-031, NUDOCS 9808120224
Download: ML031050080 (10)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

August 18, 1998

NRC INFORMATION NOTICE 98-31:

FIRE PROTECTION SYSTEM DESIGN DEFICIENCIES

AND COMMON-MODE FLOODING OF EMERGENCY

CORE COOLING SYSTEM ROOMS AT

WASHINGTON NUCLEAR PROJECT UNIT 2

Addressees

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

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

from the reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) Is Issuing this Information notice to alert

addressees to a rupture of a fire water system valve, due to a water hammer, In a fire main

vertical riser at Washington Nuclear Project Unit 2 (WNP-2) that flooded two emergency core

cooling system (ECCS) equipment rooms. 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 June 17, 1998, WNP-2 was in Mode 4 (cold shutdown) and preparations were underway for

a plant startup. At approximately 1:45 p.m., multiple fire alarms were received In the control

room coincident with three main fire pumps automatically starting and several loud water

hammer noises being heard throughout the plant. The water hammer caused a fire protection

isolation valve (FP-V-29D) to rupture In the fire protection system riser In the northeast stairwell

of the reactor building. Water from the stairwell entered residual heat removal (RHR) pump

room C through a watertight door that had not been adequately secured and began rapidly

flooding the room. A reactor drains system valve (FDR-V-609) located In a line connecting the

sumps of the RHR C and low-pressure core spray (LPCS) pump rooms failed to dose as

designed and allowed water to flow Into the LPCS pump room. The flood water completely

submerged the RHR C pump and motor and the Division 11 keepfill pump, which serves RHR B

and C and is also located In the room. Water in the LPCS pump room rose to a level just below

the pump motor and also completely submerged the minimum flow valve and the Division I

keepfill pump, which serves both the LPCS and RHR A trains.

808 12E0 98-052 180012 gal~~

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' JIN 98-31 August 18, 1998 To isolate the flooding, plant operators secured the fire pumps, which Impaired the normal fire

suppression capability of the station. On the basis of these events, the plant operators

declared a notification of unusual event (NOUE) and activated the plant emergency response

organization. As a compensatory measure for the loss of the normal fire suppression capability, the nearby Hanford fire department dispatched emergency equipment to the site.

Subsequently, the plant staff placed the fire suppression system In an alternate configuration, which was less susceptible to water hammer and terminated the unusual event.

Iti scu s sio n

I.

Fire Protection System Design and Operation

The fire protection system at WNP-2 consists of two diesel-driven and two electric- driven fire pumps. The two electric pumps and one of the diesel pumps have a capacity

of 2000 GPM each and draw a supply from the circulating water basin. The remaining

diesel-driven pump has a capacity of 2500 GPM and Is supplied by a 400,000-gallon

embankment-supported Fabritank (i.e., bladder). The fire pumps are normally in

standby and the system pressure Is maintained at approximately 150 psig by a

220-GPM jockey pump. The system Is arranged such that the pumps supply a main

header, which, In tum, supplies various yard hydrant Isolation valves and building

standpipes. The risers in the reactor building are the high points of the fire main system

at WNP-2 and rise approximately 180 feet above the main yard loop.

Additionally, the design also includes a number of PRE-ACTION systems, which are not

normally filled with water. Upon actuation of the associated detector(s) for a given PRE-

ACTION system, the PRE-ACTION system valves will open and allow water to flow from

the main header Into the associated piping. Some of the plant PRE-ACTION systems

are activated by Ionization-type detectors, whereas other PRE-ACTION systems rely on

thermal detectors for actuation. However, the sprinkler heads associated with the

downstream piping are not actuated during a PRE-ACTION unless the thermal-fusible

links are melted on the individual heads, thereby completing the flow path for the fire

water.

This event was Initiated by the actuation of fire detectors during cutting and grinding

activities, which were taking place In the diesel generator building. The fire detectors, sensing the smoke from the maintenance activities, activated a fire protection PRE-

ACTION station, which caused the associated PRE-ACTION valves to open and fill the

normally dry sprinkler line header. (A second PRE-ACTION station also actuated due to

sympathetic effects.) However, no actuation of the associated sprinklers occurred since

they are ultimately Initiated by thermal-fusible links. The depressurization of the fire

water system during the filling of the PRE-ACTION lines caused significant voiding In the

upper portions of the reactor building vertical fire main risers and generated an auto- start signal for all four main fire water pumps to start on low system pressure. Three of

the pumps started Immediately, and the fourth pump began a 30-second time delay

sequence for starting. The concurrent operation of the three pumps resulted In a rapid

reflood of the reactor building risers and collapsed the void that had been created in the

northeast stairwell riser. This sequence of events caused a significant water hammer

~'JIN 98-31 August 18, 1998 that ruptured a 12-inch, cast-iron, fire protection system Isolation valve that was located

in the stairwell riser. The licensee determined that the design of the fire protection

system was Inadequate in that the system Is configured such that destructive water

hammer forces are generated during anticipated transients when the system is in a

normal lineup. Specifically, the significant voiding caused by the PRE-ACTION

actuation, coupled with the simultaneous starting of the main fire pumps and the

unfavorable geometry of the reactor building riser and its associated supports, contributed to the severity of this event.

Common-Mode Flooding Considerations

During the event, approximately 163,000 gallons of water were Introduced into the

northeast stairwell, RHR C, and the LPCS pump rooms. Additionally, some minor

leakage of flood water occurred between the LPCS pump room and the vestibule

separating that room from the adjacent high pressure core spray pump room. Water

also leaked from the RHR C pump room Into the adjacent reactor core Isolation cooling

room through a double watertight door arrangement that separates those two rooms.

With respect to the flooding of the ECCS rooms, It was determined that the door to the

RHR C pump room was left In an unsecured condition sometime before the event. A

review of a door alarm printout from the common alarm station indicated that the door

had changed state several minutes before the fire protection system rupture. With the

door In an unsecured or open condition, an unrestricted pathway existed for flood water

to flow from the stairwell Into the RHR C pump room. However, It was noted that the

watertight doors for the northeast stairwell access to the RHR C pump room and the

LPCS pump room were not designed or Installed to prevent flooding from the stairwell

from entering the associated pump rooms (the doors were designed to seal from Inside

the rooms). Thus, even If the door had been secured, water would have entered the

pump rooms, albeit at a much slower rate. The licensee's flooding analyses had

assumed that a stairwell flood would eventually render the LPCS and RHR C systems

Inoperable and that operator actions to start the RHR A and B trains would be taken

before the loss of the keepfill pumps for those systems that are located in the LPCS and

RHR C pump rooms.

The floor drains for the RHR C pump room and the LPCS pump room drain to the same

sump. A single isolation valve, FDR-V-609, located In the drain line, Is designed to

Isolate the LPCS pump room drain from the RHR C pump room drain In the event that

the sump Is overfilled. This nonsafety-related Isolation valve is air operated via a four

way shuttle valve and accumulator. During normal operation the Isolation valve Is

opened and closed by supplying or removing air pressure to the four way shuttle valve

which ports the air to the isolation valve operator and the accumulator. The system Is

designed such that the isolation valve should fall closed on a loss of air supply pressure.

This air supply Is controlled (on/off) by a solenoid operated valve, FDR-SPV-609 which

Is upstream of the four way shuttle valve. The licensee believes that during the flooding

event the solenoid operated valve failed to fully close which resulted in a reduction of

pressure in the accumulator (resulting in a lower force to drive the sump Isolation valve

closed) and a failure to vent the air supply line to atmosphere (higher pressure In the

IN 98-31 August 18, 1998 supply line opposes closing of the sump isolation valve). As a result, the sump isolation

valve failed to close automatically when the sump reached Its high level trip point. This

allowed water from the RHR C pump room to flow through the 3-inch sump cross- connect piping and into the LPCS pump room flooding it to a level just below the pump

motor. Plant operators were unable to close the isolation valve manually from the

control room. The licensee's preliminary failure analysis Indicated that the solenoid

operated valve, an ASCO model #WJNP831654E, likely failed to operate due to age

hardening of the Buna-N diaphragm. The licensee believes that this diaphragm has not

been replaced since initial plant construction. The licensee was continuing its failure

analysis at the time this Information Notice was Issued.

Ill

Licensee Corrective Actions

As an Immediate corrective action, the licensee pumped the water from the flooded

areas. The fire protection system was returned to a functional (but degraded) status by

isolating the ruptured valve and returning the PRE-ACTION system to Its normal

condition. Subsequently, the licensee repaired or replaced all affected components.

The ruptured 12-inch cast-iron valve was replaced with a cast-steel valve. As interim

corrective actions, the licensee has established a nitrogen bubble at the top of both the

fire water system risers In the reactor building to provide a cushioning effect, and is

maintaining two fire pumps In continuous operation in order to avoid the significant

voiding expected during postulated PRE-ACTION scenarios. The licensee briefed NRC

management on the corrective actions at a public meeting In the Region IV offices in

Arlington, Texas, on July 2, 1998. The licensee committed to long-term corrective

actions, which included reviews to determine if the flooding analysis in the final safety

analysis report is adequate and whether the floor drain valves and door seals meet

design requirements. Additionally, the licensee is reviewing potential design changes

for the fire protection system to eliminate the susceptibility to water hammer. The

licensee restarted the unit on July 3, 1998.

An NRC augmented inspection team (AIT) was on site from June 17 to 23, 1998. The

results of the AIT were presented at a public exit meeting on site on July 8, 1998, and

were documented in NRC Inspection Report 50-397/98-16, which was Issued on

July 17, 1998. Preliminary Notification of Occurrence PNO-IV-98-026, which described

this event, was issued on June 18, 1998, updated on June 19, 1998, and updated again

on June 23, 1998.

-IN

98-31 August 18, 1998 This information notice requires no specific action or written response. However, licensees are

expected to review the information provided to determine whether similar system vulnerabilities

exist at their facilities. Additionally, recipients are reminded that they are required by 10 CFR

50.65 to take industry-wide operating experience (including the information presented in NRC

information notices) into consideration, where practical, 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 Office of Nuclear Reactor

Regulation (NRR) Project Manager.

Jack W. Roe, Acting Director°

Division of Reactor Project Management

Office of Nuclear Reactor Regulation

Technical contacts:

Jeffrey Shackelford, Region IV

817-860-8144 E-mail: jls2@nrc.gov

Phillip Qualls, NRR

301-415-1849 E-mail: pmq@nrc.gov

Charles Petrone, NRR

301-415-1027 E-mail cdp@nrc.gov

Attachment: List of Recently Published Information Notices

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Attachment

IN 98-31

August 18, 1998

Page 1 of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issuance

Issued to

98-30

98-29

Effect of the Year 200 Computer

Problem on NRC Licensees and

Certificate Holders

8/12/98

All material and fuel cycle

licensees and certificate holders

Predicted increase In Fuel Rod

Cladding Oxidation

Development of Systematic

Sample Plan for Operator

Licensing Examinations

Steam Generator Tube End

Cracking

813198

8/3/98

7124198

98-28

All holders of operating licenses

for nuclear power reactors, except

those licensees who have

permanently ceased operations

and have certified that fuel has

been permanently removed from

the reactor vessel.

All holders of operating licenses

for nuclear power plants

All holders of operating licenses

for pressurized-water reactors

except those who have

permanently ceased operation

and have certified that fuel has

been permanently removed for

the reactor vessel

98-27

96-48, Sup. 1 Motor-Operated Valve

Performance Issues

7/24198

All holders of operating licenses

for nuclear power reactors except

those who have permanently

ceased operation and have

certified that fuel has been

permanently removed from the

reactor vessel.

01 = Operating License

CP = Construction Permit

IN 98-31 August 18, 1998 This information notice requires no specific action or written response. However, licensees are

expected to review the information provided to determine whether similar system vulnerabilities

exist at their facilities. Additionally, recipients are reminded that they are required by 10 CFR

50.65 to take industry-wide operating experience (including the information presented in NRC

information notices) into consideration, where practical, 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 Office of Nuclear Reactor

Regulation (NRR) Project Manager.

orig Is/'d by C.l. Grimes

for

Jack W. Roe, Acting Director

Division of Reactor Project Management

Office of Nuclear Reactor Regulation

Technical contacts:

Jeffrey Shackelford, Region IV

817-860-8144 E-mail: jIs2@nrc.gov

Phillip Qualls, NRR

301-415-1849 E-mail: pmq@nrc.gov

Charles Petrone, NRR

301-415-1027 E-mail cdp@nrc.gov

Attachment: List of Recently Published Information Notices

DOCUMENT NAME: G:\\CDP\\WNP-IN.003

• SEE PREVIOUS CONCURRENCE

To receive a copy of this document, Indicate in the box C=Copy wlo attachment/enclosure E=Copy with attachment/enclosure N = No copy

OFFICE

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IN 98- xx

Julyxx, 19 Page 5 This information notice requires no specific action or written response. Ho ever, licensees are

expected to review the information provided to determine whether similar ystem vulnerabilities

exist at their facilities. Additionally, recipients are reminded that they

e required by 10 CFR

50.65 to take industry-wide operating experience (including the info

tion presented in NRC

information notices) into consideration, where practical, when settin goals and performing

periodic evaluations. If you have any questions about the inform

on in this notice, please

contact one of the technical contacts listed below or the approprite Office of Nuclear Reactor

Regulation (NRR) Project Manager.

Jack W. Ro , Acting Director

Division of

eactor Project Management

Office o0'Nuclear Reactor Regulation

Technical contacts:

Jeffrey Shackelford, Region IV

817-860-8144

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E-mail: jIs2@nrc.9ov

Phillip QuaRls, RR

301-415-184, E-mail: pm#@nrc.gov

Charles fetrone, NRR

301-41 ,-1027 E-ma cdp@nrc.gov

Attachment: List of Rece tly Published Information Notices

DOCUMENT NAME: G CDP\\WNP-IN.003

• SEE PREVIOUS CO CURRENCE

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IN 98-xx

July xx, 1998 This information notice requires no specific action or written response. However, licensees

should review the information provided to determine whether similar system vulnerabilities exist

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Jack W. Roe, Acting

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Office of Nuclea

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Technical contacts:

Jeffrey Shackelford, Region

817-860-8144 E-mail: jls2@nrc.gov

Phillip Qualls, NRR

301-415-1849 E-mail: pmq@ c.gov

Charles Pe one, NRR

301-415-

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Attachment: List of Rece y Published Information Notices

DOCUMENT NAME:

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IN 98-xx

July xx, 1998 This information notice requires no specific action or written response. However, licensees

should review the information provided to determine whether similar system vulnerabilities exist

at their facilities. Additionally, recipients are reminded that they are required by 10 CFR 50.65 to take industry-wide operating experience (including the information presented in NRC

information notices) into consideration, where practical, 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 Office of Nuclear Reactor

Regulation (NRR) Project Manager.

Jack W. Roe, Acting Director

Division of Reactor Project Management

Office of Nuclear Reactor Regulation

Technical contacts:

Jeffrey Shackelford, Region IV

817-860-8144 E-mail: jls2@nrc.gov

Phillip Qualls, NRR

301-415-1849 E-mail: pmq@nrc.gov

Charles Petrone, NRR

301-415-1027 E-mail cdp@nrc.gov

Attachment: List of Recently Published Information

DOCUMENT NAME: G:\\CDP\\WNP-IN.00%

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