Information Notice 1986-16, Failures to Identify Containment Leakage Due to Inadequate Local Testing of BWR Vacuum Relief System Valves

Failures to Identify Containment Leakage Due to Inadequate Local Testing of BWR Vacuum Relief System Valves

ML031220600
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	03/11/1986
From:	Jordan E NRC/IE
To:
References
IN-86-016, NUDOCS 8603050397
Download: ML031220600 (4)

- -ma SSINS No.: 6835 un I'sIN 86-16 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, DC 20555 March 11, 1986 IE INFORMATION NOTICE NO. 86-16: FAILURES TO IDENTIFY CONTAINMENT LEAKAGE

DUE TO INADEQUATE LOCAL TESTING OF BWR

VACUUM RELIEF SYSTEM VALVES

Addressees

All nuclear power reactor facilities holding an operating license (OL) or a

construction permit (CP).

Purpose

This notice is to alert recipients to a potentially significant problem involving

the failure to conduct adequate local leak rate tests of containment isolation

valves. It is expected that recipients will review this information for appli- cability to their facilities and consider actions, if appropriate, to preclude

a similar problem occurring at their facilities. However, suggestions contained

in this notice do not constitute NRC requirements; therefore, no specific action

or written response is required.

Past Related Correspondence:

IE Circular 77-11, "Leakage of Containment Isolation Valves with Resilient Seals"

September 6, 1977. Information Notice 79-26, "Break of Containment Integrity",

November 5, 1977. Information Notice 85-71, "Containment Integrated Leak Rate

Tests", August 22, 1985.

Description of Circumstances

During containment integrated leak rate testing, three plants had excessive

leakage associated with the torus-to-reactor-building vacuum breaker valves.

In all of these cases, the leakage was not detected by the local leak rate test

procedure because the valves were not tested with pressure applied in the

direction assumed for an accident.

Browns Ferry 2

Browns Ferry Unit 2 conducted a containment integrated leak rate test in

February 1983 that failed because of an excessive leak rate of about twice the

allowable limit of 1.5 percent per day (0.75La). The leakage path was found to

be through a flange seal on a valve in the torus-to-reactor-building vacuum

breaker system. This valve (designated FCV 64-20) is a butterfly valve bolted

8603050397

IN 86-16 March 11, 1986 into an 18-inch line connecting directly to the torus. The leakage through the

flange seal was reduced to an acceptable rate by tightening flange bolts.

Local leak rate testing, which is required to be performed every 2 years, is

done by applying pressure between FCV 64-20 and a flapper-type check valve that

is located on the reactor building side of the butterfly valve. However, the

leaking flange was on the torus side of FCV 64-20. Consequently, the valve

flange was not included in the local testing, but was tested only during the

integrated testing which is done every 3 to 4 years.

Peach Bottom 2

Peach Bottom Unit 2 conducted a containment integrated leak rate test in June

1985 that produced an excessive leak rate of about three times the allowable

limit of 0.375 percent per day. Most of the leakage was found to be going

through the stem seal of valve AO-2502B, an air-operated butterfly valve located

adjacent to the torus in the vacuum breaker line. An apparently successful

local leak rate test performed on this valve prior to the integrated test had

failed to detect the leakage. Local leak rate testing is done by applying

pressure between valve AO-2502B and the check valve located between the reactor

building and this valve. However, the valve stem for AO-2502B is located on the

torus side of the valve and, as in the Browns Ferry case, this leak path was not

subject to the local leak rate test pressure.

Duane Arnold

During a containment integrated leak rate test at Duane Arnold in July 1985, difficulty was experienced in establishing the test pressure. The problem was

found to be caused by leakage through a hole left by a plug that was missing

from the body of isolation valve CV4305. This valve was part of the

torus-to-reactor-building vacuum breaker system and was located on the torus

side of the vacuum breaker line. The plug had evidently been removed during

maintenance conducted on the valve during the same outage as the integrated

test. An apparently successful local leak rate test, conducted on the valve

after the maintenance, had failed to detect the hole. This failure was due to

the fact that the hole was located on the torus side of the valve disc, and

the test pressure had been applied to the other side of the valve.

Discussion:

NRC regulations (10 CFR 50, Appendix J, Section III.C.1) require that local leak

rate test pressure be applied in the same direction as that which would exist

when the valve would be required to perform its safety function, unless it can

be determined that the results from tests for a pressure applied in a different

direction will provide equivalent or more conservative results. Many facilities

experience problems in applying this rule because of the difficulty of applying

a local test pressure for large isolation valves connected directly to primary

containments. After the Browns Ferry test failure, TVA identified 14 containment

isolation valve flanges on each of the Browns Ferry units that were not being

tested under the local leak rate test procedures then in use. After the Peach

Bottom test, two valves on Unit 2 and five valves on Unit 3 were found to be

oriented so that the valve stems were not being subjected to local leak rate

test pressure.

IN 86-16 March 11, 1986 There are modifications and test techniques that can be applied to cause the

local leak rate test to produce "equivalent or more conservative results." For

example, at Browns Ferry, TVA is committed to solving the valve flange problem

by installing double seals (gaskets) on the problem flanges. Local leak rate

test pressure can be applied between the seals to produce a local test that can

be considered equivalent to or more conservative than internal pressurization.

This technique may also be used on valve stems that are designed to permit

double seals. In some situations valve stem seals may be included in the

normally pressurized boundary by turning the valve around without reducing the

effectiveness of the valve. In some cases special test devices such as a blank

flange may be used to seal the line inboard of the inner isolation valve.

No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional

Administrator of the appropriate regional office or this office.

Edwar Hi. Jordan, Director

Divisi'n of Emergency Preparedness

and Engineering Response

Office of Inspection and Enforcement

Don Kirkpatrick, IE

(301) 492-4510

Attachment: List of Recently Issued IE Information Notices

1 - -

Attachment 1 IN 86-16 March 11, 1986 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue Issued to

86-15 Loss Of Offsite Power Caused 3/10/86 All power reactor

By Problems In Fiber Optics facilities holding

Systems an OL or CP

86-14 PWR Auxiliary Feedwater Pump 3/10/86 All power reactor

Turbine Control Problems facilities holding

an OL or CP

86-13 Standby Liquid Control 2/21/86 All BWR facilities

System Squib Valves Failure holding an OL or CP

To Fire

86-12 Target Rock Two-Stage SRV 2/25/86 All power reactor

Setpoint Drift facilities holding

an OL or CP

86-11 Inadequate Service Water 2/25/86 All power reactor

Protection Against Core Melt facilities holding

Frequency an OL or CP

84-69 Operation Of Emergency Diesel 2/24/86 All power reactor

Sup. 1 Generators facilities holding

an OL or CP

86-10 Safety Parameter Display 2/13/86 All power reactor

System Malfunctions facilities holding

an OL or CP

86-09 Failure Of Check And Stop 2/3/86 All power reactor

Check Valves Subjected To facilities holding

Low Flow Conditions an OL or CP

86-08 Licensee Event Report (LER) 2/3/86 All power reactor

Format Modification facilities holding

an OL or CP

86-07 Lack Of Detailed Instruction 2/3/86 All power reactor

And Inadequate Observance Of facilities holding

Precautions During Maintenance an OL or CP

And Testing Of Diesel Generator

Woodward Governors

OL = Operating License

CP = Construction Permit