Potential for Surveillance Testing to Fail to Detect an Inoperable Main Steam Isolation ValveML031070034 |
Person / Time |
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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, 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, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Issue date: |
02/01/1994 |
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From: |
Grimes B Office of Nuclear Reactor Regulation |
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To: |
|
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References |
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IN-94-008, NUDOCS 9401260242 |
Download: ML031070034 (10) |
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Similar Documents at 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, 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, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Category:NRC Information Notice
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Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555 February 1, 1994 NRC INFORMATION NOTICE 94-08: POTENTIAL FOR SURVEILLANCE TESTING TO FAIL
TO DETECT AN INOPERABLE MAIN STEAM ISOLATION
VALVE
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
PurDose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice (IN)to alert addressees to a potential for surveillance testing to
fail to detect that a main steam isolation valve is mechanically bound and
will not close. 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.
DescriDtion of Circumstances
On April 19, 1993, while performing maintenance to repair a presumed faulty
limit switch on a main steam isolation valve, the licensee for the River Bend
Station (River Bend) found that the valve was mechanically bound and would not
close as required. The valve is a 24-inch-diameter (nominal), spring and
pneumatic closing, pneumatic-opening, internally balanced, poppet-type globe
valve manufactured by the Atwood & Morrill Company Inc. Plant operators had
previously performed partial stroke surveillance testing of the valve on
February 27 and April 1, 1993, but did not detect that the valve would not
close. The licensee later determined that the testing failed to detect that
the valve was inoperable because the test did not adequately consider the
design of the valve and the positioning of the limit switch arm in relation to
the valve poppet travel.
The licensee determined that the valve would not close because improper
clearances between the valve poppet and the valve body had caused excessive
wear of the guide ribs and resulted in the valve poppet becoming mechanically
bound. The excessive wear may have been avoided had the licensee installed an
anti-rotation modification recommended by the manufacturer in 1989.
Subsequent to this event, the manufacturer reported the failure to close to
the NRC under Part 21 to Title 10 of the Code of Federal Regulations and
informed affected licensees of the failure mechanism and recommended actions
to prevent recurrence. NRC Inspection Report 50-458/93-18 and Licensee Event
Report.93-006 provide additional details on the valve failure.
(9401260242) TJ'O
02 w> 44M'j J7JD#V&//c/S
YJ Kx/ IN 94-08 February 1, 1994 Discussion
The original design positioning of the limit switches was such that, during
partial stroke testing, the limit switches could be actuated and indicate
movement of the main valve poppet even though the valve poppet had not
actually moved. Thus, a failure of the valve to properly stroke may go
undetected by partial stroke testing.
Under normal operation, the valve stem travels 28 centimeters [11 inches] to
fully stroke in either direction. As the valve strokes open, the first
2.5 centimeters [1 inch) of stem travel moves an internal poppet which opens
an equalizing port allowing the pressure on both sides of the main poppet to
equalize. During the remainder of the open stroke, the internal poppet lifts
the main poppet and retracts it to the fully open position. During a partial
stroke test in the close direction, as the stem (and the internal poppet)
begins to close, the main poppet also begins to close because of gravity.
However, during the event, with the main poppet stuck in the open position, the stem travelled about 2.5 centimeters [I inch] and stopped when the
internal poppet seated in the equalizing port.
There are three limit switches on the valve that are of concern in this event.
The first two switches provide a safety-related signal to the reactor
protection system that the valve is 92 percent open. The third switch sends a
nonsafety-related signal to position indicating lights in the control room
indicating that the valve is 90 percent open. The licensee had set the
90-percent-open limit switch such that stem movement of about 2.8 centimeters
[1.1 inch] was required to actuate the switch and indicate that the main valve
poppet had moved to the 90-percent-open position. However, because the limit
switches are set with a + 2 percent tolerance, actual stem travel to actuate
the 90-percent-open limit switch may be only 2.25 centimeters [0.88 inch]. In
a worst-case scenario both the 92- and the 90-percent-open limit switches
could be actuated without the main valve poppet moving.
During the partial stroke testing conducted on February 7 and April 1, the
first two limit switches (92-percent-open indication) actuated, the third
limit switch (90-percent-open indication) did not actuate. Although the
procedural step called for receipt of the 90-percent-open indication, the
operators did not declare the valve inoperable because the first two limit
switches had actuated and they assumed that the third limit switch (nonsafety)
had failed. Later, on April 17, during maintenance on the presumed faulty
limit switch, the licensee found that the main valve poppet was mechanically
bound and that the valve would not close.
The licensee for River Bend changed the third limit switch setting to actuate
at 85 percent of the open position to ensure that its actuation during partial
stroke testing would give positive indication of poppet movement. Pending
further evaluation of these valves during the next refueling outage, the
licensee is performing full stroke testing of the valves on a quarterly basis
and intends to install the anti-rotation modification recommended by the
vendor to prevent recurrence of the excessive wear of the valve guides. Also, operations personnel have been trained on the operation and function of the
limit switches.
--- IN 94-08 February 1, 1994 The valve described in this notice is used in safety-related applications at
nuclear facilities. One such application is as a main steam isolation valve.
At River Bend and most domestic boiling water reactors, there are two main
steam isolation valves for each main steam line; an inboard valve, located
inside the drywell, and an outboard valve, located Just outside the primary
containment. After a design-basis accident, these valves are required to
close and remain closed for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Should these valves fail to close, offsite dose limits could be exceeded. A similar failure of a main steam
isolation valve to close had occurred at a foreign boiling water reactor.
The potential for limit switch positioning to adversely affect surveillance
test accuracy may exist for valves other than that described in this notice.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the persons listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager. by
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: Thomas F. Westerman, RIV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, RIV William M. McNeill, RIV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
v-" IN 94-08 February 1, 1994 The valve described in this notice is used in safety-related applications at
nuclear facilities. One such application is as a main steam isolation valve.
At River Bend and most domestic boiling water reactors, there are two main
steam isolation valves for each main steam line; an inboard valve, located
inside the drywell, and an outboard valve, located Just outside the primary
containment. After a design-basis accident, these valves are required to
close and remain closed for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Should these valves fail to close, offsite dose limits could be exceeded. A similar failure of a main steam
isolation valve to close had occurred at a foreign boiling water reactor.
The potential for limit switch positioning to adversely affect surveillance
test accuracy may exist for valves other than that described in this notice.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the persons listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager. Original t1gnad b
Brian K. Grimes
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: Thomas F. Westerman, RIV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, RIV William M. McNeill, RIV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
OFFICE RPB:ADM EMEB:DE:NRR C/EMEB:DE:NRR PDIV-2 NAME RSanders* PCampbell* JNorberg* EBaker*
DATE 11/18/93 12/21/93 12/22/93 12/27/93 REGION IV REGION IV C/ES:RIV DD/DRS:RIV D/DRS:RIV
WMcNeill* DLoveless* TWesterman* AHowell* SCollins*
12/27/93 01/03/94 01/03/94 01/03/94 01/04/94 of
OIP/NA/Per _ OGCB:DORS:NRR C/OGCB:DORS:NRR D
KHenderson* JBirmingham* GHMarcus* Wg9
01/06/94 01/10/94 01/10/94 01/Z7/94 OFFICIAL DOCUMENT NAME: 94-08. IN
IN 94-xx
January 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
one of the persons 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 contacts:
Thomas F. Westerman, Region IV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, Region IV William M. McNeill, Region IV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
OFFICE RPB:ADM EMEB:DE:NRR lC/EMEB:DE:NRR PDIV-2 NAME RSanders* PCampbell* JNorberg* EBaker*
DATE 11/18/93 12/21/93 l12/22/93 12/27/93 REGION IV [REGION IV C/ES:RIV DD/DRS:RIV lD/DRS:RIV
WMcNeill* DLoveless* TWesterman* AHowell* SCollins*
12/27/93 01/03/94 01/03/94 01/03/94 01/04/94 OIP/NA/Per
KHenderson*
01/06/94 OGCB:DORS:NRR
JBirmingham*
01/10/94 IC/OGCB:DORS:NRR
GHMarcus
01/10/94 D/DORS:NRR (
BKGrimes
01/ /94 I
OFFICIAL DOCUMENT NAME: ATWOODIN.JLBA
IN 94-xx
January 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
one of the persons 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 contacts:
Thomas F. Westerman, Region IV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, Region IV William M. McNeill, Region IV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
OFFICE RPB:ADM EMEB:DE:NRR C/EMEB:DE:NRR PDIV-2 NAME RSanders* PCampbell* JNorberg* EBaker*
DATE 11/18/93 12/21/93 12/22/93 12/27/93 REGIONY, EG10 YVI C/ES:RI ee DD/DR ' D/DRS:R ' ,,20d
'7--
WM6N Tte 'ste a nAH &i' SColl I i 12+
12/27/93 01/03/94 01/03/94 01/03/94 01/04/94 OIP/NA er OGCB: DORS:NK, C/OGCB:DORS:NRR D/DORS:NRR
KHen6'ri¶"'1 JBI rminghar GHMarcus BKGrimes
01/06/94 101//0/94 01/ /94 01/ /94 OFFICIAL DOCUMENT NAME: ATWOODIN.JLB
IN 93-xx
January x, 1993 stroke testing would give positive indication of poppet movement. Pending
further evaluation of these valves during the next refueling outage, the
licensee is performing full stroke testing of the valves on a quarterly basis
and is considering installing the anti-rotation modification recommended by
the vendor to prevent recurrence of the excessive wear of the valve guides.
Also, operations personnel have been trained on the operation and function of
the limit switches.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the persons 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 contacts:
Thomas F. Westerman, Region IV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, Region IV William M. McNeill, Region IV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
OFFICE EMEB:DE:NRR C100AWR PDIV-2 NAME PCampbel
__JN___rg__ _ I
DATE 12/AI/93 12/1ZI93 12/27/93 I
OFFICE }jGIONI REGION IV [REGION IV F
NAME
_McNeil f' DLoveless TWesterman
DATE 12A_7/93____ 12/ /93 12/ /93 12/ /93 OFFICE RPB:ADM OGCB:DORS:NRR C/OGCB:DORS:NRR D/DORS:NRR
NAME RSanders* JBirmingham GHMarcus BKGrimes
DATE 11/18/93 12/ /93 12/ /93 12/ /93 A.F IC A _U U L l NAME: .' R LI._.^
L ,A~wu v r. . .J_.
M FIGIAL DOUMEN AMUMOIN.JLB
IN 93-xx
January x, 1993 stroke testing would give positive indication of poppet movement. Pending
further evaluation of these valves during the next refueling outage, the
licensee is performing full stroke testing of the valves on a quarterly basis
and is considering installing the anti-rotation modification recommended by
the vendor to prevent recurrence of the excessive wear of the valve guides.
Also, operations personnel have been trained on the operation and function of
the limit switches.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the persons 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 contacts:
Thomas F. Westerman, Region IV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, Region IV William M. McNeill, Region IV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
OFFICE EMEB:DE:NRR C/EMEB:DE:NRR PDIV-2 NAME PCampbell JNorberg I
DATE 12/ /93 12/ /93 I 12/ /93 OFFICE REGION IV REGION IV REGION IV
NAME WMcNeill DLoveless TWesterman
DATE 12/ /93 12/ /93 12/ /93 12/ /93 OFFICE R1B:Ap L OGCB:DORS:NRR C/OGCB:DORS:NRR D/DORS:NRR
NAME RSanders JBirmingham GHMarcus BKGrimes
DATE 11/18/93 12/ /93 12/ /93 12/ /93 OFFICIAL DOCUMENT NAME: ATWOODIN.JLH
Atta -ement, IN 94-08 February 1, 1994 Page I of 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information Date of
Notice No. Subject Issuance Issued to
93-26, Grease Solidification 01/31/94 All holders of OLs or CPs
Supp. 1 Causes Molded-Case for nuclear power reactors.
Circuit Breaker Failure
to Close
94-07 Solubility Criteria for 01/28/94 All byproduct material and
Liquid Effluent Releases fuel cycle licensees with
to Sanitary Sewerage Under the exception of licensees
the Revised 10 CFR Part 20 authorized solely for
sealed sources.
94-06 Potential Failure of 01/28/94 All holders of OLs or CPs
Long-Term Emergency for boiling water reactors.
Nitrogen Supply for the
Automatic Depressurization
System Valves
93-85, Problems with X-Relays 01/20/94 All holders of OLs or CPs
Rev. 1 in DB- and DHP-Type for nuclear power reactors.
Circuit Breakers Manu- factured by Westinghouse
94-05 Potential Failure of 01/19/94 All holders of OLs or CPs
Steam Generator Tubes for pressurized water
with Kinetically Welded reactors (PWRs).
Sleeves
94-04 Digital Integrated 01/14/94 All NRC licensees except
Circuit Sockets with licensed operators.
Intermittent Contact
94-03 Deficiencies Identified 01/11/94 All holders of OLs or CPs
during Service Water System for nuclear power reactors.
Operational Performance
Inspections
94-02 Inoperability of General 01/07/94 All holders of OLs or CPs
Electric Magne-Blast for nuclear power reactors.
Breaker Because of Mis- alignment of Close-Latch
Spring
OL = Operating License
CP = Construction Permit
K-' xJ IN94-08 February 1, 1994 The valve described in this notice is used in safety-related applications at
nuclear facilities. One such application is as a main steam isolation valve.
At River Bend and most domestic boiling water reactors, there are two main
steam isolation valves for each main steam line; an inboard valve, located
inside the drywell, and an outboard valve, located Just outside the primary
containment. After a design-basis accident, these valves are required to
close and remain closed for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Should these valves fail to close, offsite dose limits could be exceeded. A similar failure of a main steam
isolation valve to close had occurred at a foreign boiling water reactor.
The potential for limit switch positioning to adversely affect surveillance
test accuracy may exist for valves other than that described in this notice.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the persons listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager. Original igned Iy
Brian K. Grimes
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: Thomas F. Westerman, RIV Patricia Campbell, NRR
(817) 860-8145 (301) 504-1311 David P. Loveless, RIV William M. McNeill, RIV
(512) 972-2507 (817) 860-8174 Attachment:
List of Recently Issued NRC Information Notices
OFFICE RPB:ADM [EMEB:DE:NRR lC/EMEB:DE:NRR l PDIV-2 NAME RSanders* PCampbell* JNorberg* EBaker*
DATE 11/18/93 12/21/93 12/22/93 12/27/93 REGION IV REGION IV C/ES:RIV DD/DRS:RIV D/DRS:RIV
WMcNeill* DLoveless* TWesterman* AHowell* SCollins*
12/27/93 01/03/94 01/03/94 01/03/94 01/04/94 do
OIP/NA/Per OGCB:DORS:NRR C/OGCB:DORS:NRR
KHenderson* JBirmingham* GHMarcus*
01/06/94 01/10/94 01/10/94 0 1/Z7/94 OFFICIAL DOCUMENT NAME: 94-08.IN
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list | - Information Notice 1994-01, Turbine Blade Failures Caused by Torsional Excitation from Electrical System Disturbance (7 January 1994)
- Information Notice 1994-02, Inoperability of General Electric Magne-Blast Breaker Because of Misalignment of Close-Latch Spring (7 January 1994)
- Information Notice 1994-03, Deficiencies Identified During Service Water System Operational Performance Inspections (11 January 1994, Topic: Biofouling)
- Information Notice 1994-04, Digital Integrated Circuit Sockets with Intermittent Contact (14 January 1994)
- Information Notice 1994-05, Potential Failure of Steam Generator Tubes with Kinetically Welded Sleeves (19 January 1994, Topic: Stress corrosion cracking)
- Information Notice 1994-06, Potential Failure of Long-Term Emergency Nitrogen Supply for the Automatic Depressurization System Valves (28 January 1994)
- Information Notice 1994-07, Solubility Criteria for Liquid Effluent Releases to Sanitary Sewerage Under the Revised 10 CFR Part 20 (28 January 1994)
- Information Notice 1994-08, Potential for Surveillance Testing to Fail to Detect an Inoperable Main Steam Isolation Valve (1 February 1994)
- Information Notice 1994-09, Release of Patients with Residual Radioactivity from Medical Treatment & Control of Areas Due to Presence of Patients Containing Radioactivity Following Implementation of Revised 10 CFR Part 20 (3 February 1994, Topic: Brachytherapy)
- Information Notice 1994-10, Failure of Motor-Operated Valve Electric Power Train Due to Sheared or Dislodged Motor Pinion Gear Key (4 February 1994)
- Information Notice 1994-11, Turbine Overspeed and Reactor Cooldown During Shutdown Evolution (8 February 1994, Topic: Overspeed trip)
- Information Notice 1994-12, Insights Gained from Resolving Generic Issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment (9 February 1994, Topic: Overspeed)
- Information Notice 1994-13, Unanticipated and Unintended Movement of Fuel Assemblies and Other Components Due to Improper Operation of Refueling Equipment (28 June 1994)
- Information Notice 1994-14, Failure to Implement Requirements for Biennial Medical Examinations and Notification to the NRC of Changes in Licensed Operator Medical Conditions (24 February 1994, Topic: Overspeed)
- Information Notice 1994-15, Radiation Exposures During an Event Involving a Fixed Nuclear Gauge (2 March 1994, Topic: Overspeed trip)
- Information Notice 1994-16, Recent Incidents Resulting in Offsite Contamination (3 March 1994, Topic: Overspeed trip)
- Information Notice 1994-17, Strontium-90 Eye Applicators: Submission of Quality Management Plan (QMP), Calibration, and Use (11 March 1994, Topic: Brachytherapy, Overspeed)
- Information Notice 1994-17, Strontium-90 Eye Applicators: Submission of Quality Management Plan (Qmp), Calibration, and Use (11 March 1994, Topic: Brachytherapy, Overspeed)
- Information Notice 1994-18, Accuracy of Motor-Operated Valve Diagnostic Equipment (Responses to Supplement 5 to Generic Letter 89-10) (16 March 1994)
- Information Notice 1994-19, Emergency Diesel Gemerator Vulnerability to Failure from Cold Fuel Oil (16 March 1994)
- Information Notice 1994-20, Common-Cause Failures Due to Inadequate Design Control and Dedication (17 March 1994)
- Information Notice 1994-21, Regulatory Requirements When No Operations Are Being Performed (18 March 1994)
- Information Notice 1994-22, Fire Endurance & Ampacity Derating Test Results for 3-Hour Fire-Rated Thermo-Lag 330-1 Fire Barriers (16 March 1994, Topic: Fire Barrier)
- Information Notice 1994-23, Guidance to Hazardous, Radioactive and Mixed Waste Generators on the Elements of a Waste Minimization Program (25 March 1994, Topic: Fire Barrier)
- Information Notice 1994-24, Inadequate Maintenance of Uninterruptible Power Supplies & Inverters (24 March 1994, Topic: Safe Shutdown, Fire Barrier)
- Information Notice 1994-25, Failure of Containment Spray Header Valve to Open Due to Excessive Pressure from Inertial Effects of Water (15 March 1994, Topic: Fire Barrier, Water hammer)
- Information Notice 1994-26, Personnel Hazards and Other Problems from Smoldering Fire-Retardant Material in the Drywell of a Boiling-Water Reactor (28 March 1994, Topic: Fire Barrier)
- Information Notice 1994-27, Facility Operating Concerns Resulting from Local Area Flooding (31 March 1994, Topic: Fire Barrier)
- Information Notice 1994-28, Potential Problems with Fire-Barrier Penetration Seals (5 April 1994, Topic: Fire Barrier, Operability Assessment)
- Information Notice 1994-29, Charging Pump Trip During a Loss-of-Coolant Event Caused by Low Suction Pressure (11 April 1994, Topic: Boric Acid)
- Information Notice 1994-30, Leaking Shutdown Cooling Isolation Valves at Cooper Nuclear Station (19 August 1994, Topic: Fire Barrier, Local Leak Rate Testing)
- Information Notice 1994-31, Potential Failure of Wilco, Lexan-Type HN-4-L Fire Hose Nozzles (14 April 1994, Topic: Hydrostatic, Stress corrosion cracking)
- Information Notice 1994-32, Revised Seismic Estimates (29 April 1994, Topic: Stress corrosion cracking, Earthquake)
- Information Notice 1994-33, Capacitor Failures in Westinghouse Eagle 21 Plant Protection Systems (9 May 1994, Topic: Stress corrosion cracking)
- Information Notice 1994-34, Thermo-LAG 330-660 Flexi-Blanket Ampacity Derating Concerns (13 May 1994, Topic: Fire Barrier, Stress corrosion cracking)
- Information Notice 1994-35, Niosh Respirator User Notices, Inadvertent Separation of the Mask-Mounted Regulator(Mmr) from the Facepiece on the Mine Safety Appliances (16 May 1994, Topic: Stress corrosion cracking)
- Information Notice 1994-35, Niosh Respirator User Notices, Inadvertent Separation of the Mask-Mounted Regulator(MMR) from the Facepiece on the Mine Safety Appliances (16 May 1994)
- Information Notice 1994-36, Undetected Accumulation of Gas in Reactor Coolant System (24 May 1994, Topic: Reactor Vessel Water Level, Stress corrosion cracking, Integrated leak rate test)
- Information Notice 1994-37, Misadministration Caused by a Bent Interstitial Needle During Brachytherapy Procedure (27 May 1994, Topic: Stress corrosion cracking, Brachytherapy)
- Information Notice 1994-38, Results of Special NRC Inspection at Dresden Nuclear Power Station, Unit 1 Following Rupture of Service Water Inside Containment (27 May 1994)
- Information Notice 1994-39, Identified Problems in Gamma Stereotactic Radiosurgery (31 May 1994)
- Information Notice 1994-40, Failure of a Rod Control Cluster Assembly to Fully Insert Following a Reactor Trip at Braidwood, Unit 2 (26 May 1994)
- Information Notice 1994-41, Problems with General Electric Type Cr124 Overload Relay Ambient Compensation (7 June 1994)
- Information Notice 1994-41, Problems with General Electric Type CR124 Overload Relay Ambient Compensation (7 June 1994)
- Information Notice 1994-42, Cracking in the Lower Region of the Core Shroud in Boiling-Water Reactors (7 June 1994, Topic: Stress corrosion cracking)
- Information Notice 1994-43, Determination of Primary-to-Secondary Steam Generator Leak Rate (10 June 1994, Topic: Grab sample)
- Information Notice 1994-44, Main Steam Isolation Valve Failure to Close on Demand Because of Inadequate Maintenance and Testing (16 June 1994)
- Information Notice 1994-44, Main Steam Isolation Valve Failure to Close on Demand because of Inadequate Maintenance and Testing (16 June 1994)
- Information Notice 1994-45, Potential Common-Mode Failure Mechanism for Large Vertical Pumps (17 June 1994, Topic: Biofouling)
- Information Notice 1994-46, Nonconservative Reactor Coolant System Leakage Calculation (20 June 1994, Topic: Unidentified leakage)
... further results |
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