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 L 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

Technical

Contact: 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 86-14 86-13 86-12 86-11 84-69 Sup. 1 86-10 86-09 86-08 86-07 Loss Of Offsite Power Caused By Problems In Fiber Optics Systems PWR Auxiliary

Feedwater

Pump Turbine Control Problems Standby Liquid Control System Squib Valves Failure To Fire Target Rock Two-Stage

SRV Setpoint Drift Inadequate

Service Water Protection

Against Core Melt Frequency 3/10/86 3/10/86 2/21/86 2/25/86 2/25/86 All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All BWR facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP Operation

Of Emergency

Diesel 2/24/86 Generators

Safety Parameter

Display 2/13/86 System Malfunctions

Failure Of Check And Stop 2/3/86 Check Valves Subjected

To Low Flow Conditions

Licensee Event Report (LER) 2/3/86 Format Modification

Lack Of Detailed Instruction

2/3/86 And Inadequate

Observance

Of Precautions

During Maintenance

And Testing Of Diesel Generator Woodward Governors OL = Operating

License CP = Construction

Permit