Information Notice 1994-30, Leaking Shutdown Cooling Isolation Valves at Cooper Nuclear Station

Leaking Shutdown Cooling Isolation Valves at Cooper Nuclear Station

ML031060492
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:	08/19/1994
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-94-030, Suppl 1, NUDOCS 9408160070
Download: ML031060492 (9)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C.

20555

August 19, 1994

NRC INFORMATION NOTICE 94-30, SUPPLEMENT 1:

LEAKING SHUTDOWN COOLING

ISOLATION VALVES AT

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 (IN) supplement to clarify the type of valve discussed in the original

notice and to alert recipients to the potential that local leak rate testing

of primary containment isolation valves with pressure applied in the direction

opposite of the accident direction may not be conservative. 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.

Background

The NRC issued IN 94-30, "Leaking Shutdown Cooling Isolation Valves at Cooper

Nuclear Station," on April 12, 1994, to alert addressees to a precursor to an

unisolable rupture of shutdown cooling piping with the potential for core

damage and release of radioactive material outside the containment.

In that

event, after receiving a high-pressure alarm for the suction piping in the

residual heat removal system, the licensee measured the leakage through the

inboard isolation valve, established a bleedoff path into the pressure

maintenance system and continued operations. About 10 months later, the

licensee disassembled the inboard and outboard valves and found cracks in the

seating surfaces of both valves.

During the preparation of IN 94-30, the staff determined that a concern may

exist regarding the method used for testing of primary containment isolation

valves. This concern is described below. In addition, the NRC has received

information from Anchor/Darling Valve Company that the description of the

valves in IN 94-30, "Anchor-Darling 20-inch nominal, double-disk, flex-wedge, gate valves," is incorrect. The correct description is, "a 20 inch, 900

class, flex wedge gate valve," manufactured by the Anchor Valve Company before

the merger that formed the Anchor/Darling Valve Company.

In IN 94-30, the NRC

staff was concerned about the actions of the licensee after the valve leakage

was identified regardless of the valve type or manufacturer.

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IN 94-30, Supplement 1 August 19, 1994

Description of Circumstances

On May 1, 1993, after evaluating the results of local leak rate testing for

several inboard primary containment isolation valves, the licensee for the

Cooper Nuclear Station (Cooper) determined that the testing methodology used

for Appendix J, Type C tests of certain flex-wedge gate valves could not be

relied on to be equivalent to or more conservative than testing the valves in

the accident direction.

The Type C tests for several such valves had been

performed with the pressure applied in the direction opposite to the accident

direction (reverse direction). The licensee began corrective actions and

notified the NRC in Licensee Event Report (LER)93-019, dated June 1, 1993.

In LER 93-019, the licensee reported that the testing of these valves in the

reverse direction was based on an incorrect interpretation of information from

the valve manufacturer. In response to a question from the licensee regarding

the acceptability of valve test methods, the manufacturer had indicated that

testing in the reverse direction would not be expected to affect test results, but an unqualified answer could not be provided. Results of recent tests by

the licensee of some of the primary containment isolation valves with pressure

applied in the accident direction, indicate that testing in the reverse

direction may not be equivalent to or more conservative than testing in the

accident direction; therefore earlier test results may not be valid.

LER 93-019 indicated that Cooper Nuclear station has 65 inboard primary

containment valves of various types and from various manufacturers.

The

licensee found that 24 of those valves were of concern. Of the 24 valves,

8 could be tested in the accident direction but, before the 1993 refueling

outage, were not.

The other 16 valves could not be tested in the accident

direction nor had testing in the reverse direction been qualified to be

equivalent or more conservative than testing in the accident direction. For

most of these valves, the licensee qualified testing in the reverse direction

to be equivalent or more conservative or modified the testing configuration to

allow testing the valve in the accident direction. For the remaining valves, the licensee requested exemptions from certain Appendix J requirements to

allow leak .rate testing in the reverse' direction.

Discussion'

Appendix J to Title 10 of the Code of Federal Regulations Part 50 requires a

program for leak testing the primary reactor containment and related systems

and components penetrating the primary containment pressure boundary.

Section III.C.1 of Appendix J specifies that, for Type C tests of valves

(local leak rate tests), the pressure shall be applied in the same direction

as that when the valve would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure

applied in a different direction will provide equivalent or more conservative

results. This requirement is intended to ensure that test leak rates are

representative of leak rates that would be experienced under actual accident

conditions. As described above, the licensee for Cooper found that, for some

of the primary containment isolation valves, testing in the reverse direction

was not equivalent to or more conservative than testing in the accident

direction.

IN 94-30, Supplement 1

August 19, 1994 Related Generic Communications

• NRC IN 86-16, "Failures to Identify Containment Leakage Due to

Inadequate Local Leak Rate Testing," March 11,

1986.

NRC IN 92-20, Inadequate Local Leak Rate Testing,' March 3, 1992.

This information notice requires no specific action or written response. If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian

s, Director Q

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:

Jim Pulsipher, NRR

(301) 504-2811 Attachment:

List of Recently Issued NRC Information Notices

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Attachment

IN 94-30, Supp. 1

August 19, 1994 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issuance

Issued to

94-59

94-58

94-57

94-56

94-55

Accelerated Dealloying of

Cast Aluminum-Bronze

Valves Caused by Micro- biologically Induced

Corrosion

Reactor Coolant Pump

Lube Oil Fire

Debris in Containment

and the Residual Heat

Removal System

Inaccuracy of Safety Valve

Set Pressure Determinations

Using Assist Devices

Problems with Copes-

Vulcan Pressurizer

Power-Operated Relief

Valves

Deficiencies Found in

Thermo-Lag Fire Barrier

Installation

Failures of General

Electric Magne-Blast

Circuit Breakers to

Latch Closed

Possible Malfunction of

Westinghouse ARD, BFD,

and NBFD Relays, and

A200 DC and DPC 250

Magnetic Contactors

08/17/94

08/16/94

08/12/94

08/11/94

08/04/94

08/04/94

08/01/94

07/29/94

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for pressurized water

reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

91-79, Supp. 1

94-54

91-45, Supp. 1

94-42, Supp. 1

Cracking in the Lower

Region of the Core Shroud

in Boiling-Water Reactors

07/19/94 All holders

for boiling

(BWRs).

of OLs or CPs

water reactors

OL = Operating License

CP = Construction Permit

V IN

'-30,

Supplement 1

August 19, 1994 This information notice requires no specific action or written response. If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

orig /s/'d by CIGrimes/for

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact: Jim Pulsipher, NRR

(301) 504-2811 Attachment:

List of Recently Issued NRC Information Notices

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DATE

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Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:

Jim Pulsipher, NRR

(301) 504-2811 Attachment:

List of Recently Issued NRC Information Notices

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the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:

Jim Pulsipher, NRR

(301) 504-2811 Attachment:

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IN 44-30, Supplement 1 August XX, 1994 This information notice requires no specific action or written response. If

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:

Jim Pulsipher, NRR

(301) 504-2811 Attachment:

List of Recently Issued NRC Information Notices

• See Previous Concurrence

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

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact:

Jim Pulsipher, NRR

(301) 504-2811 Attachment:

List of Recently Issued NRC Information Notices

• See Previous Concurrence

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