ML20054A083
| ML20054A083 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Dresden |
| Issue date: | 09/22/1981 |
| From: | ORGANIZATION FOR ECONOMIC COOPERATION & DEVELOPMENT |
| To: | |
| Shared Package | |
| ML19240B432 | List: |
| References | |
| FOIA-81-380 NUDOCS 8204150271 | |
| Download: ML20054A083 (5) | |
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.r. u r. w SEP 2 21981 No. IRS 59 RESTRICTED DIFFUSION RESTREINTE Date of Receipt Date de Rdception 8th September 1981 Name of nuclear power station Nom de la centrale Dresden 2 (lfnited States')
Date of incident Date de l' incident 5th March 1981
.t Type of reactor Type de r6acteur BWR Authorized electrical power Niveau de puissanc'e diectrique 794 MWe autorisd First commercial operation l
Date de mise en service July 1970 l
Included is the extraction from " Power Reactor Events" i
USNRC, Vol. 3, No. 3, 1981, with the approval of the IRS Coordinator in the United States.
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8204150271 811109 PDR FOIA CONNOR 81-380 PDR 4765 r
EMPLOYEE RADIATION EXPOSURE DUE TO LOW WATER LEVEL IN REACTOR VESSEL On March 5,1981 at Dresden Unit 2,"* a. licensee contractor employee received a whole body exposure of 21 rems while guiding a crane in the removal of temporary concrete shielding from inside the reactor vessel.
The water level beneath the shielding had apparently dropped and was 'not detected by a control room conitoring device.
Radioactive components exposed by the low water level set off alarms and the area was evacuated.
Although there were other workers in the area at the time of the incident, only the one worker received an over-exposure.
The NRC limit for radiation workers is 3 rems per calendar quarter.
The overexposure occurred while preparations were being made to remove a two-segment-concrete shield plug from the defueled reactor cavity following completion of feedwater sparger replacement work during an outage begun on January 2.
The maintenance foreman checked with the control room and was advised that the indicated water level'in the cavity was 57" above instrument "0" for the instrument in use.
This corresponds to at least 20" above the top of the core shroud.
It was subsequently determined that the water level in the reference leg of the instrument being used to monitor reactor vessel water level was actually down between 27"' and 37".
This resulted in a corresponding error in the indicated vessel water level of between 20" and 30".
That is, the actual water level was between 8"' and 18" below the top of the core shroud.
(See Figure 1.),
The two segments of the concrete shield plug rested on a circular I-beam placed on the core shroud, and also functioned as a working plationn during ' tor replacement of the feedwater spargers.
On the evening of March 5, a contrac.
mi11 wright entered the reactor cavity to assure that the first plug, which was' being raised by an overhead crane, did not bind on guide pins or strike the internal components.
When the segment was raised about three and one-half feet, an area radiation monitor 'on the refuel floor sounded and personnel in the area were evacuated.
The millwright had been exposed to high radiation levels, estimated to be in the range of 500 to 1000 P/hr, emanating from the unshielded vessel internals.
Emergency processing of his film badge and finger thermoluminescent dosimeter (TLD) showed readings of 21.2 rems and 20.8 rems, respectively.
Personnel monkoring devices worn by. other personnel present in the area or who had entered the reactor cavity earlier in the evening showed no other exposures in excess of regulatory limits.
The overexposed individual underwent a medical examination on March 6.
The results of a chromosome aberration study were consistent with the badge and TLD data, but a blood count taken <an the indi-vidual showed no abnormalities.
Water level in the reactor had been maintained using two devices.
One was a temporary float device to be used at.the side of the shield plug, which was equipped with brackets on which the float device.was installed.
It was cap-able of monitoring.-a range of 22" with high and low alarm points that could be set at any level aTong the detector range, and.would read out the level and sound an alarm on the reactor building refuel floor.
The alarm setpoints had arbitrarily been set at,20% and 80% of full scale.
"A 794 MWe BWR located 9 miles east o'f Morris, Illinois, and operated by I
Commonwealth Edison Coesany.
I
For water level indication during.the outages, the level identified as the top of the active fuel was established as instrument "0."
On this basis, the top of the shroud was about 37".
The bottom of the ~ plug, resting on' a 14" ring girder, was at about 51", and the top of the plug was close to the 81" level.
Safe end work was to be performed during the outage on the lower instrument nozzle, which was at 74" (Figure 1).
It was decided that water level should be maintained as far as practicable below the nozzles because of the welding to be done, since during past outages there had been problems in maintaining close control of water level. A 10" level below the nozzle, 60" 5", was set.
The float indicator had been installed and disconnected early in the outage.
Alarms had been experienced frequently because the lower alarm setpoint (62.4") was at or close to the desired water level.
The operations engineer responsible did not know how the instrument worked and was not aware that the alarm setpoints were adjustable.
The location of the float indicator also could have been adjusted by equipping the device with brackets,.,
The other water level monitoring device used was a W/R-GEMac, an initrument that senses the difference in height of two water columns and converts the difference to an" electrical signal that is transmitted to 'its appropriate indicating device showing level.
In this case, the range was converted to 0"-200" and was continuously recorded in the control room.
The two water
-O columns (or levels) sensed were the reference' and variable legs. The refer.
ence leg is normally full of w'ater up to its penetration in the reactor vessel (high), and the variable leg is the actual level of water in the reactor vessel (low)'.
The difference in heights (pressures) of the variable and the
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reference legi is sensed by a device that compares the two pressures (differ-ential pressure) which is converted to the electrical signal that is transmitted to the control room.
Since the reference leg is assumed to remain constant by engineering design, the level in the variable les (reactor vessel level) is the parameter that changes, and thus a change in the vessel level is shown on c
the recorder.
A calibration check of the'W/R-GEMac performed in February showed that the 6:.trument was accurate.
Total reliance was then placed on the device to pro-vide water level indication, since the shielding installed prohibited visual checks of the vessel water level during sparger work, even though certain personnel had expressed some concern over a lack of redundancy in water level indication.
During the outage, the reactor vessel penetration for the reference leg was cut near the reactor vessel for nondestructive testing, and no provision was made for keeping it full.
Normally, during operation it is kept full by condensing steam from the reactor vessel.
Following the incident,. personnel checked the W/R-GEMac instrumentation and, by backfilling the reference leg, determined that the water level in it was down 27".
This correspondingly caused the instru=ent to indicate that water level in the vessel was 27" higher than it actually was.
Another refilling of the reference leg revealed an additional 10" discrepancy applied to conditions at the time of the incident.
In either case, some of the highly radioactive vessel. internals were not covered by water when the shield plug was raised on March 5...
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k' While the feedwater sparger work was being done, the control rod drives (CRDs) 3 were also being exchanged. With each CRD withdrawal, varying amounts of water were lost from the vessel.
The lost water was being made up several times per j
shift, up to the required indicat.ed level, but th'ese daily variations appar-ently masked the loss of water from the reference leg until the shield plug
'l was pulled.
Between February 10 and 28, however, no additions were made even though CRD changes continued.
Since the record of water additions was not I
trended, the lack of water additions to the reactor vessel went unnoticed.
(
The specific cause of the water loss from.the reference leg has not been determined.
One theory is that evaporation occurred, accelerated by heat 2
generated during welding activities on a tap which was cut to perform safe end -
3 work.
This is unlikely, however, since the heliarc welding using an Argon 1
purge did not transmit heat beyond a few inches of the weld location.
A j
second possibility is that incorrect valving operations occurred during cali-
,j brations of other instrumentation tied into this reference leg, although there was no indication of an unaccounted for step or rapid change in indicated I
vessel level that would indicate a sudden drop in the reference leg level had
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occurred.
A third possibility is that water was lost through a slow leak in 1
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one of the valves in the instrumentation system.
Only one cup of water from the reference leg could result in a 27" error in indicated vessel water level.
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The licensee plans to upgrade he'alth physics requirements covering feedwater j
sparger replacement so that, if a remote detector cannot be used, a radiation y
control technician with a portable survey meter will accompany personnel j
working inside the vessel.
The com.and and control func, tion of the shift <
supervisor on the off-shifts will also be strengthened.
On March 5, the H
operations engineer believed that he had left sufficient directions to ensure q
that the, shift foreman would be present and involved during the critical d
activities on the shift during which the overexposure occurred. That direc-H tion did not reach the personnel actually lifting the shield plugs, however, a
and the shift foreman and engineer were unaware that the shield plug was being j
lifted.
In addition, it is now clear that no single water level instrument should be relied on during maintenance programs similar to the feedwater sparger program, and that detailed health physiEs~ procedures should exist and be employed'for critical operations.3 R
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SEP 2 21981 No. IRS 60 RESTRICTED DIFFUSlON RESTREINTE Date of Receipt 15th September 1981 Date de R6ception Name of nuclear power station Generic Problem Nom de la centrale (United States)
Date of incident 18th August 1981 Date de l' incident (Date of report)
Type of reactor PWRs (Block valves of Type de r6acteur power-operated relief valves)
Authorized electrical power Niveau de puissance 61ectrique autorisd First commercial operation Date de mise en service l
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T 4765
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SSIN No. 6820 Accession'No.:
8011040283 IEB 81-02, Supp. 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
. WASHINGTON, D.C."
20555 August l'8, 1981 IE BULLETIN NG. 81-02 SUPPLEMENT:
FAILURE OF GATE TYPE VALVES TO CLOSE AGAINST DIFFERENTIAL PRESSURE Descriotion of Circunstances:
IE Bulletin No. 81-02, " Failure of Gate Type Valves to Close Against Differential Pressure," identified several gate type valves that had been shown by tests and/or analyses by Westinghouse to have.a potential for not closing against differential pressure.
As a part of its ongoing analysis program, Westinghouse Electro-Mechanical Division (W-EMD).has applied the analytical methods developed for valves discussed'~in IE Bulletin 81-02 to the remaining motor operated gate valves that they manufacture.
These analyses predict that closure problems could also be anticipated with 6, 8, 10, 12, 14, 16, and 18-in'ch. nominal size valves in addition to 3-and 4-inch low pressure valves.
Thus, the entire line of W-EMD manufactured motor operated gate valves has the potential for not clqsing against differential pressure.
Westinghouse has indicated to NRC that they have notified.all of their domestic -
nuclear customers of this problem.
When.the valves were provided as original scope of supply, they also identified the recommended corrective action necessary to assure valve closure under the system service conditions that their records show the valves will experience.
Where the valves were provided as spares or replacements, they indicated the threshold differential pressure across the valves above which closure could not be assured under the utility order equip-ment specification conditions.
A list of power reactor facilities that Westinghouse believes to have the potentially affected valves is given in Table 1.
However, this list, as sell as Westinghouse's notifications, does not take into consideration the fact that the valves may have.been transferred between facilities.
In addition, the Westinghouse determinations of operability and corrective action do not take into consideration the fact that the valves may have been transferred between systems or that the system service conditions may have changed through design evolution.
It is therefore essential that all facilities verify the presence or absence of W-EMD manufactured" motor-operated gate valves and also verify their ability to close under the current and/or intended service conditions.
Actions To Be Taken by Licensees:
1.
Within 30 days of the issuance date of this bulletin supplement, ascertain whether any WyEMD manufactured motor operated gate valves have been VQ$5
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Page 2 of 4 installed, or are maintained as spares for installation, in safety-related systems where they are required to close against a differential pressure.
2.
If no valves, other than those reported in response to IE Bulletin 81-02, are, identified, report this to be the, case.
No further action is required.
3.
If any valves are identified as being installed, verify that they are capable of closing under their current limiting normal and post' accident service conditions.
If such cannot be shown, take corrective action on s
these affected valves,and evaluate. the effect that failure to close under any condition requiring closure would have on system (s) operability pursuant to the facility technical specifications for dontinued operation.
e 4.
If any valves are identified as spares, verify that they are capable of closing under their intended limiting normal and post accident service I
conditions.
If such cannot be shown, either modify the affected valves so that they are qualified for the intended service or obtain qualified replacements prior to installation.
3 5.
Within 45 days of the issuance dat'e' of this bulletin supplement, submit a report to NRC listing the affected valves. identified in safety related systems, their service or planned service, and the maximum differential pressure at which they would be required to close, the safety consequences of the valves failing to close, the corrective action taken or planned, and the schedule for completing the corrective action.
Actions To Be Taken by Construction Permit Holders:
~
1.
Ascertain whether any W-EMD manufactured motor-operated gates valves are or will be installed, or maintained as spares for installation, in safety-related systems where they are. required to close against a differential-pressure.
2.
If no valves, other than those reported in response to IE Bulletin 81-02, are identified, report this to be the case.
No further action is required.
3.
If any valves are identified, verify that they are capable of closing under their intended limiting normal and post accident service conditions.
If such cannot be shown, either modify the affected valves so that they are qualified for the intended service or obtain qualified replacements prior to startup.
4.
Within 90 days of the issuance dat'e of this bulletin supplement, submit a report to NRC listing the affected valves identified for use in safety-related systems, their planned service, the maximum differential pressure at which they Nould be required to close, the safety consequences of the valves failing to close, the corrective action taken or planned, and the schedule for completing the corrective action.
For those cases in which reports have already been submitted in accordance with the Technical Specification, 10 CFR Parts 21 and/or 50.55(e), this information need not be resubmitted.
Rather,' licensees or construction permit holders should reference this earlier report and submit only the additional information requested above.
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IEB 81-02, Supp. 1 August 18, 1981,
-o Page 3 of 4 Reports, signed under oath or affirmation under the provisions of Section 182a of the Atemic Energy Act of 1954, shall be submitted to the Director of the appropriate NRC Regional Office and a copy shall be forwarded to the Director of the NRC Office of Inspection and Enforcemsnt, Washington, D.C.
20555.
If you need additional information regarding this matter, please contact the appropriate NRC Regional Office.
This request for information was approved by OMB under blanket clearance number 3150-0012 that expires December 31, 1981.
Comments on burden and duplication should be directed to the Office of Management and Budget, Reports Management Room 3208, New Executive Office Building, Washington, D.C.
20503.
Attachments:
1.
Table 1, Partial List of Plants With Valves Manufactured by W-EMD 2.
Recently issued IE Bulletins g e 6 @
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IEB 81-02,~ Suppo 1 August 18, 1981 '
O TABLE 1.
PARTIAL LIST OF PLANTS WITH VALVES MANUFACTURED BY W-EMD Valves Supplied as Valves Supplied as Spares or Replacements Original Scope of Supply Arkansas Nuclear One 1 Beaver Valley 2 Beaver Valley 1 Braidwood 1, 2 Callaway 1, 2 Byron 1, 2 Catawba 1, 2 Diablo Canyon Callaway 1, 2 Comanche Peak 1, 2 Farley Shear.on Harris 1, 2, 3 Indian Point 2 Marble Hill 1, 2 Midland 1,.2
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South Texas 1, 2 Seabrook 1, 2 Dyster Creek Prairie Island 1,2 Summer 1 St. Lucie 2 Vogtle;1, 2 San Onofre 1, 2, 3 Watts Bar 1, 2 Summer 1 Wolf Creek 1 Surry 2
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IEB 81-02, Supp. 1
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August 18, 1981 '
RECENTLY ISSUED IE BULLETINS
' Bulletin No.
Subject Date Issued Issued To 81-03 Flow Blockage of Cooling 4/10/81 All nuclear power Water to Safety System facilities with Components by Corbicula an OL or CP Sp. (Asiatic Clam) and Mytilus Sp. (Mussel) 81-02 Failure of Gate Type Valves 3/9/81 All power reactor to Close Against Differential facilities with an Pressure OL or CP 81-01 Surveillance of Mechanical 3/5/81 All power reactor
. Rev. 1 Snubbers facilities with an OL & specified facilities with CP 80-17, Failure of Control Rods 2/13/81 All BWR facilities Supplement 5 to Insert _During a Scram with OL or CP 81-01 Surveillance of 1/27/81 All power reactor Mechanical Snubbers facilities with!DL
& to specified facilities with CP,
80-25 Y Operating Problems with 12/19/80 All BWR facilities Target Rock Safety-Relief with OL & specified Vaives at BWRs near-term OL BWR facilities & all BWRs 2
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with a CP e.;
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Supplement 4
' Failure of Control 'Ro'ds
'. ~.7112/18/80 To specified BWRs to 80-17 to Insert D0 ring'a Scrani - ' 'i-
~l' d 'i BWRs with a CP-
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80-24 Prevention oV Damig'e- ~ ~ "
'11/21/80 All power reactor i
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Due to Water Leakage facilities with Inside Containment OL or CP (October 17, 1980 Indian Point 2 Event) 80-23 FaHures of Solenoid 11/14/80 All power reactor Valves Manufactured by facilities with Valcor Engineering OL or CP Corporation OL = Operating License CP = Construction Permit g
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J SEP 2 21981 No. IRS 60.2 RESTRICTED DIFFUSlON ItESIREINTE Date of Receipt Date de R6ception 15th September 1981 Name of nuclear power station Generic Problem Nom de la centrale (United States)
Date of incident 9th April 1981 Date de l' incident (Date of report)
. e.
Type of reactor PWRs (Block valves of Type de r6acteur power-operated relief valves)
Authorized electrical power Niveau de puissance diectrique autorisd First commercial operation Date de mise en service l
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4765
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[ '11N No. 6E20 M
's..ccession No.:
8011040283
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IEB 81-02 UNITED STATES HI NUCLEAR REGULATORY COMMISSION lf 0FFICE OF INSPECTION AND ENFORCEMENT
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WASHINGTON, D.C.
20555 April 9',1981 IE Bulletin No. 81-02:
FAILURE OF. GATE TYPE VALVES TO CLOSE AGAINST DIFFERENTIAL f PRESSURE 9
Descriotion of Circumstances:
l As a part of its pressurized water reactor (PVR) Safety and Relief Valve Testing tj Program, the Electric Power Research Institute (EPRI) conducted limited testing 4
of a number of valves used on PWRs as power-cperated. relief valve (PORV) a isolation or block valve's.
These tests indicate a number of cases in which rf
. certain of these valves failed to fully close under conditions that approximated
'i those of their intended service (i.e., saturated steam at approximately 2,400 psi' I J The valves 'that failed to fully close are gate type motor-operated valves that j
may be used in various safety-related applications in addition to-PORV block N
valves.
i Backoround on EPRI Testina:
P
.h 4
The proposed full-scale qualification testing of PORV block valves, with a' lj completion date of July 1,1982, was first provided to the utilities in a H
September 5, 1980, draft of NUREG-0737.
The item was formally issued, with J
Commission approval, in NUREG-0737 on October 31, 1980.
fa The block valve qualification testing was proposed in NijREG-0737 primarily as an U
additional means of reducing the number of challenges to the emergency core-
[
cooling system and the safety valves during plant operation.
l'
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In anticipating a request for PWR block valve testing, EPRI de'cided to make provisions for the ingta11ation of block valves be' tween the test steam source 1
and the test PORV in July 1980 at the Marshall test facility.
The Marshall test facility is a full-flow steam. test facility.pwned by Duke Power Company.
4 Test PORVs had been carefully selected, with close coordi. nation between EPRI, its consultants and PWR utilities, to assure that PORVs representative of those
]l in service'or intended for service would be tested.
However, for the block valves that have been' tested concurrently, this selection process was not
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followed because an NRC block valve test program had not been formulated.
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Therefore, seven readily available valves were obtained and tested by EPRI, primarily to obtain some general baseline information on block valve closure l
capability.
g 4
For the block valves that were tested, EPRI had not established, at least at the-
- 4 time of testing, the population of plants, either operating or under construction, a
that might have a valve of the type needed for testing.
In addition, it should be noted that the test conditions used at Marshall to date were only those that ha were determined to be applicable for steam testing of PORVs.
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EB S1-02 april 9, 1981 i
Page 2 of 4 These test conditions were selected after review by EPRI, utilities, and PWR NSSS vendors. NRC staff also reviewed and concurred with the test conditions.'
To date, there has been no similar specific determination by EPRI or the NRC staff as to th'e relevance of the Marshall, block' valve test conditions to the conditions -in any specific PWR plant under which a block valve should be able to close to isolate a stuck-open PORV.
To date, EPRI has tested a total of seven PORV block valves, all at the Marshall !
facility.
During these tests, the following valves failed to fully close during the EPRI PORV block valve testing:
1.
Westinchouse Electro-Mechanical Division (W-EMD) 3-inch Valves - These valves, which are manufactured by W-EMD, can be identified by the yoke-mounted nameplates.that are stamped " WESTINGHOUSE" and include " VALVE IDENT." and " VALVE I.D." numbers given in Table 1.
Supplemental analyses and water testing, performed by W-EMD, determined that a 4-inch valve also would not close fully and therefore is included in this bulletin. The nameplate'. data on this valve are given in Table 1.
These analyses and tests aisc determined the thresh:1d differential pressure across the valves above which closure cannot be assured.
Thes'e val.ues are-given in Table 1.
A list of power reactor facilities believed to have the affected valves is given in Table 2.
It is our understanding that W-EMD has notified these facilities of the failure.of these valves to ful_ly close.-
l 2.
Boro-Warner' Nuclear Valve Division (BW-NVD) 3-inch 1500-pound Motor-Operated Gate Valves - These valves can be identified by BW-NVD part numbers 75460, 77910, and 79190.
Supplemental testing to detennine threshold differential pressures for less severe service has yet to be completed. A list of power reactor facilities believed to have the affected valves is given in Table 3.
BW-NVD has submitted a 10 CFR Part 21 report in which they indicated that they have notified these facilities of the failure of these valves to fully close.
(Note:
Similar valves with BW-NVD part numbers 74380 and 74380-1 have been modified,' retested, and demonstrated to close under test conditions.' As a result, they are not included in this bulletin.)
3.
Anchor Darlina 3-inch 1540 pound Double-Dis'c Valve - This valv he first of a, series of specially designed valves, has been modified, retested, and demonstrated to close unde,r test conditions ~.
The remaining valves will be similarily modified during manufacture.
As a result, they are not included in this bulletin.
kt must be cautioned that Tables 2 and 3 may not be complete.
For example, the staff is aware of one power reactor facility that obtained affected valves from anotherinventory: 'For this reason, this bulletin is applicable to all power reactor facilities with an operating license or construction permit.
The tests and analyses performed to date raise doubts as to the ability of the affected valves to close under less severe service conditions.' These valves have also been supplied for utilization in a nu=ber of safety-related 3
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EB 81-02 1
q.pril 9, 1981 Page 3 of 4 applications.
In the case of the W-EMD valves, they are also provided as spares or replacements through direct sales from the manuf acturer.
For this reason, this bulletin is applicable to the affected' valves that are required to close with a, differential pre.ssure across them,in safety-related systems or as PORV block valves..
The responsibility for notification and corrective actions based on adverse test results continues to lie with the utilities and vendors in the industry.
NRC will continue to monitor the progress of the qualification program. All adverse test data will continue to be evaluated on a case-by-case basis.
NRC staff will take appropriate action, if necessary, to assure that the necessary corrective actions are made in a timely manner.
Actions to be Taken by Licensees:
1.
Within 30 days of the issuance date of this bulletin, ascertain whether any of the affected valves have been installed, or 'are maintained as spares for installation, where they are required to close with a differential pressure across them in safety-related systems or as PORV block valves.
The differ-ential pressures of concern include the fol, lowing:.
a.
For the W-EMD manufactured valves, values in excess of the threshold values iii Table 1.
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b.
For the BW-NVD valves',- any value.
2.
If no affected valves are identified, report this to be the case and
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ignore the items below.
3.
If any affected valves are identified as being installed, take corrective action and evaluate the effect that failure to close under any conditi'on requiring closure would have on system (s) operability pursuant to the facility technical specifications for continued operation.
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4.
If any affected valves are identified as spares, either modify the valves so that they are qualified for the intended service or obtain qualified replacements prior to installation.
5.
Within 45 days of the issuance date of this bulletin, submit a report to NRC listing the affected v'alves identified, their service or planned service, the maximum differential pressure at which they vould be required to close, the safety consequences of the valve's failure to close, the corrective action taken or planned, and the schedule for completing the corrective action.
Actions to be Taken by Construction Permit Holders:
1.
Ascertain whether any of the affected valves are or will be installed or maintained as spares for installation where they are required to close
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\\_.spril 9, 1981 Page 4 of 4 with a differential pressure act,oss them in safety related systems or as PORV block valves. The differential pressures of concern include the following:
For the W-EMD manufactured valv'es, values in excess of the threshold
~
a.
v'alues Oi Table 1.
b.
For the BW-NVD valves, any value.
2.
If no affected valves are identified, report this to be the case and ignore
~
the items below.
3.
If any affected valves are identified, either modify the valves so that they are qualified for the intended service or obtain qualified replacements prior to startup.
4.
Within 90 days of the issuance date of'this bulle' tin, submit a repor't to NRC listing the affected valves identified, their planned service, the maximum differential pressure at which they would be required to close, the safety consequences of the valve's failure to q1ose, the corrective action taken or planned, and the schedule for comp 1.eting the corrective action.
4 For those cases in which reports have already been submitted in accordancs with the Technical Specification,10 CFR Parts 21 and/or 50.55(e), this information need not be resubmitted.
Rather, licensees or construction permit holders should reference this earlier report and submit only the additional information requested above.
Reports, signed under oath or affirmation under the provisions of Section 182a of the Atomic Energy Act of 1954, shall be submitted to the Director of the -
appropriate NRC Regional Office and a copy shall be forwarded to the Director of the NRC Office of Inspection and Enforcement, Washington, D.C.
20555.
If you need additional information regarding this matter, please contact the appropriate NRC Regional Office.
~
This request for information was approved by GAO under blanket clearance number R0072 that expires November 30, 1983.
Comments on burden and duplication should be directed to Office'of Managefnent and Budget, Room 3201, New Executive Office Building, Washington, D.C.
20503.
Attachments:
~
1.
Table 1 - Identification of W-EMD Manufactured Valves and Differential Pressure Limits for Operation 2.
Table 2 - Partial List of Plants With Affected Valves Manufactured by W-EMD 3.
Table 3 - Partial List of Plants WIth Affected Valves Manufactured by BW-NVD 4.
Recently issued IE Bulletins J
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TABLE 1.
IDENTIFICATION OF }/-EMD MANUFACTURED VALVES AND DIFFERENTIAL PRESSURE LIMITS FOR VALVE OPERATION Nominal W-EMD Valve Ilodel
" VALVE A***l Size (in.)
Reference IDENT."*
" VALVE I.D."**
(psid)'
3 3GPB8 03000GM88 3GM58 or 3GM78 or 3GM88 1500 3GM88 03002GM88 3G,M58 or 3GM78 or 3GM88 1500 3GM99 03001GM99 3GM58 or 3GM78 or 3GMS8 750 4
4GM88 04000GM88 4GM78 or 4GM88 750l 4GM88
~~ 04002GM88 4GM78 or 4GM88 750 4
4GM87 04000GM87 4GM77 750 4GM87 04002GM87 4GM77 750 This number is found on the yoke ~ mounted nameplate and occupies the first ni.ne positions of a 24 position number.
It'is used in evaluating the functional'AP requirements.
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'b This number. is found on the yoke-mounted nameplate and occupies the first three positions of a six position number.
Valves sold as spares or replacements may not contain this number.
% u-Pressure below which valve will close (as shipped).
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Notes: A " position" may contain more than one character.
The three position
" VALVE I.D." number consists of five digits in the three positions; for example, 3 GM 78.
l A1T nameplates have " VALVE IDENT." numbers, but those sold as spares or replacements may not. have " VALVE-I.D." numbers.
The " VALVE IDENT."
. number includes the manufacturer's model reference, and the " VALVE I.D." ' number is a reference to the valve system application.
The
" VALVE I.D." number also appears on Westinghouse valve indexes and.
system flow diagrams.
There is no reference to the " VALVE IDENT."
nu=ber on these indexes or flow diagrams.
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TABLE 2.
PARTIAL LIST OF PLANTS.WITH
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AFFECTED VALVES MANUFACTURED BY _W-EMD
" VALVE IDENT." Humber 04000GMS8 04002GMS8 s
03000GM88 04000GM87 03002GM88 03001GM99 04002GMS7 Plant Operating plants (supplied as spares or replacements except as noted):
Beaver Valley 1 X
Connecticut Yankee X
Farley 1, 2 X*
Indian Point 2' X
Kewaunee X
Horth Anna 1, 2 X
Oconee 1, 2, 3 "X
X San Onofre 1 X..
Surry 1, 2 X
X X
-4 Zion 1, 2 X
Nonopera'ing plants (supplied as original scope'of supply except as noted):
t Beaver Valley 2 X
X Braidkood 1, 2 X
X
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Byron 1, 2 X
.X Callaway 1, 2 -
X X
Comanche Peak 1, 2 X
X Harris 1, 2', 3, 4 X
X-Jamesport 1,'.2 X
X.._
Marble. Hill 1, 2 X
X San Onofre 2, 3 X**
Seabrook 1, 2 X
X South Texas 1, 2 X
Summer X
X Vogtle 1, 2 X
X Watts Bar 1, 2 X'
X Wolf Creek X
X
- Transferred from inventory at another plant.
- Spares or replacements.
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\\pril 9, 1981 IEB 81-02 TABLE 3.
PARTIAL LIST OF PLANTS WITH AFFEC.TED. VALVES MANUFACTURED BY BW-tND Plant NVD-P/N Arkansas Nuclear One, Unit '2 75460 Bellefonte 79190 Palo Verde 77910
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('..ttachment 4 M.EB 81-02 s.-
April 9, 1981 RECENTLY ISSUED IE BULLETINS Bulletin No.
Subject Date Issued Issued To 81-01
. Surveillance of Mechanical-3/5/81 All power reactor Rev. 1 Snubbers facilities with an OL & specified.
facilities with a CP 80-17, Failure of Control Rods 2/13/81 All BWR facilities Supplement 5 to Insert During a Scram with an OL or CP 81-01 Surveillance of 1/27/81
-- All power reactor Mechanical Snubbers facilities with an OL
& to specified facilities with a CP
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80-25 Operating Problems with 12/19/80 All BWR facilities
' Target Rock Safety-Relief
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with an OL & specifies Valves at BWRs near tenn OL & all BWI facilities with a CP Supplement 4 Failure of Control Rods 12/18/80 To specified BWRs to 80-17 to Insert During a Scram with an OL & All at a.BWR BWRs with a.CP 80-24 Prevention of Damage
- 11/21/80.
AllpoweSreactor
- Due to Water Leakage facilities with an Inside Contairment OL or CP (October 17, 1980 Indian Point 2
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, :. V 80-23 7 Failures of Solenoid 11/14/80 All power reactor Valves Manufactured by facilitie's with an Valcor Engineering OL or CP Corporation 80-22
, Automation Industries, 9/11/80 All radiography Model 200-520-008 Sealed-licensees Source Connectors 8b-21
- Valve yokes supplied by 11/6/80 All light water Malcolm Foundry Company, Inc.
reactor facilities with an OL or CP s
OL = Operating License CP = Construction Permit E..~k :.*..
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