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

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{{#Wiki_filter:--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
{{#Wiki_filter:--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==
==Addressees==
:All nuclear power reactor facilities holding an operating license (OL) or aconstruction permit (CP).
:
All nuclear power reactor facilities
 
holding an operating
 
license (OL) or a construction
 
permit (CP).


==Purpose==
==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.
: 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


==Description of Circumstances==
Observance
: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 pressure.


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
Of Precautions


===Technical Contact:===
===During Maintenance===
Don Kirkpatrick, IE(301) 492-4510Attachment: List of Recently Issued IE Information Notices
And Testing Of Diesel Generator Woodward Governors OL = Operating


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
License CP = Construction


}}
Permit}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Revision as of 13:01, 31 August 2018

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)
v • d • e


--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


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