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.: 6835un I'sIN 86-16UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF INSPECTION AND ENFORCEMENTWASHINGTON, DC 20555March 11, 1986IE INFORMATION NOTICE NO. 86-16: FAILURES TO IDENTIFY CONTAINMENT LEAKAGEDUE TO INADEQUATE LOCAL TESTING OF BWRVACUUM RELIEF SYSTEM VALVES

Addressees

All nuclear power reactor facilities holding an operating license (OL) or aconstruction permit (CP).

Purpose

This notice is to alert recipients to a potentially significant problem involvingthe failure to conduct adequate local leak rate tests of containment isolationvalves. It is expected that recipients will review this information for appli-cability to their facilities and consider actions, if appropriate, to precludea similar problem occurring at their facilities. However, suggestions containedin this notice do not constitute NRC requirements; therefore, no specific actionor 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 RateTests", August 22, 1985.

Description of Circumstances

During containment integrated leak rate testing, three plants had excessiveleakage 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 testprocedure because the valves were not tested with pressure applied in thedirection assumed for an accident.Browns Ferry 2Browns Ferry Unit 2 conducted a containment integrated leak rate test inFebruary 1983 that failed because of an excessive leak rate of about twice theallowable limit of 1.5 percent per day (0.75La). The leakage path was found tobe through a flange seal on a valve in the torus-to-reactor-building vacuumbreaker system. This valve (designated FCV 64-20) is a butterfly valve bolted8603050397 IN 86-16March 11, 1986 into an 18-inch line connecting directly to the torus. The leakage through theflange seal was reduced to an acceptable rate by tightening flange bolts.Local leak rate testing, which is required to be performed every 2 years, isdone by applying pressure between FCV 64-20 and a flapper-type check valve thatis located on the reactor building side of the butterfly valve. However, theleaking flange was on the torus side of FCV 64-20. Consequently, the valveflange was not included in the local testing, but was tested only during theintegrated testing which is done every 3 to 4 years.Peach Bottom 2Peach Bottom Unit 2 conducted a containment integrated leak rate test in June1985 that produced an excessive leak rate of about three times the allowablelimit of 0.375 percent per day. Most of the leakage was found to be goingthrough the stem seal of valve AO-2502B, an air-operated butterfly valve locatedadjacent to the torus in the vacuum breaker line. An apparently successfullocal leak rate test performed on this valve prior to the integrated test hadfailed to detect the leakage. Local leak rate testing is done by applyingpressure between valve AO-2502B and the check valve located between the reactorbuilding and this valve. However, the valve stem for AO-2502B is located on thetorus side of the valve and, as in the Browns Ferry case, this leak path was notsubject to the local leak rate test pressure.Duane ArnoldDuring a containment integrated leak rate test at Duane Arnold in July 1985,difficulty was experienced in establishing the test pressure. The problem wasfound to be caused by leakage through a hole left by a plug that was missingfrom the body of isolation valve CV4305. This valve was part of thetorus-to-reactor-building vacuum breaker system and was located on the torusside of the vacuum breaker line. The plug had evidently been removed duringmaintenance conducted on the valve during the same outage as the integratedtest. An apparently successful local leak rate test, conducted on the valveafter the maintenance, had failed to detect the hole. This failure was due tothe fact that the hole was located on the torus side of the valve disc, andthe 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 leakrate test pressure be applied in the same direction as that which would existwhen the valve would be required to perform its safety function, unless it canbe determined that the results from tests for a pressure applied in a differentdirection will provide equivalent or more conservative results. Many facilitiesexperience problems in applying this rule because of the difficulty of applyinga local test pressure for large isolation valves connected directly to primarycontainments. After the Browns Ferry test failure, TVA identified 14 containmentisolation valve flanges on each of the Browns Ferry units that were not beingtested under the local leak rate test procedures then in use. After the PeachBottom test, two valves on Unit 2 and five valves on Unit 3 were found to beoriented so that the valve stems were not being subjected to local leak ratetest pressur IN 86-16March 11, 1986 There are modifications and test techniques that can be applied to cause thelocal leak rate test to produce "equivalent or more conservative results." Forexample, at Browns Ferry, TVA is committed to solving the valve flange problemby installing double seals (gaskets) on the problem flanges. Local leak ratetest pressure can be applied between the seals to produce a local test that canbe considered equivalent to or more conservative than internal pressurization.This technique may also be used on valve stems that are designed to permitdouble seals. In some situations valve stem seals may be included in thenormally pressurized boundary by turning the valve around without reducing theeffectiveness of the valve. In some cases special test devices such as a blankflange 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 RegionalAdministrator of the appropriate regional office or this office.Edwar Hi. Jordan, DirectorDivisi'n of Emergency Preparednessand Engineering ResponseOffice of Inspection and Enforcement

Technical Contact:

Don Kirkpatrick, IE(301) 492-4510

Attachment:

List of Recently Issued IE Information Notices 1 --Attachment 1IN 86-16March 11, 1986LIST OF RECENTLY ISSUEDIE INFORMATION NOTICESInformation Date ofNotice No. Subject Issue Issued to86-1586-1486-1386-1286-1184-69Sup. 186-1086-0986-0886-07Loss Of Offsite Power CausedBy Problems In Fiber OpticsSystemsPWR Auxiliary Feedwater PumpTurbine Control ProblemsStandby Liquid ControlSystem Squib Valves FailureTo FireTarget Rock Two-Stage SRVSetpoint DriftInadequate Service WaterProtection Against Core MeltFrequency3/10/863/10/862/21/862/25/862/25/86All power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll BWR facilitiesholding an OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPOperation Of Emergency Diesel 2/24/86GeneratorsSafety Parameter Display 2/13/86System MalfunctionsFailure Of Check And Stop 2/3/86Check Valves Subjected ToLow Flow ConditionsLicensee Event Report (LER) 2/3/86Format ModificationLack Of Detailed Instruction 2/3/86And Inadequate Observance OfPrecautions During MaintenanceAnd Testing Of Diesel GeneratorWoodward GovernorsOL = Operating LicenseCP = Construction Permit