ML20005B714
| ML20005B714 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 08/19/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| IEB-81-02, IEB-81-2, NUDOCS 8109010612 | |
| Download: ML20005B714 (2) | |
Text
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AUGUST 1 9 1981 Docket No. 50-395 l
MDIORANDUM FOR: Atomic Safety and Licensing Board for Virgil C. Summer Nuclear Station, Unit No.1 i
FRON:
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing NRR
SUBJECT:
BOARD NOTIFICATION - FAILURE OF PORY BLOCK VALVE TO CLOSE DURING TESTS CONDUCTED BY EPRI (BN No. 81-Z1 )
As a part of its pressurized water reactor (PWR) safety and relief valve testing program, the Eletric Power Research Institute (EPRI) conducted limited testing of a number of valves used on PWRs as power operated relief valve (PORV) isolation block valves. These tests indicate a number of cases in which certain of these valves failed to fully close during conditions that approximated those of their intended service. A more complete description of the testing program and the actions to be taken by licensees and applicants is contained in Attachment 1 to this memorandum, IE Bulletin 81-02, " Failure of Gate-Type Valves to Close Against Differential Pressure."
The purpose of this menorandum is to explain the relevance of these block valve failures to the Virgil C. Summer Nuclear Station Unit No.1.
As discussed in IE Bulletin 81-02, the Virgil C. Summer Nuclear Station, Unit i
No.1, design incorporates block valves of a type which failed during the testing program. The applicant's response to IE Bulletin 91-02 is included as Attachment 2 to this memorandum. Additional clarification was requeted by NRR and the applicant's response is included as Attachment 3 to this memorandum.
In sumary, the affected valves have been modified and retested with acceptable resul ts. The same modification was made to the block valves used in the Virgil C.
Summer Nuclear Station, Unit No.1.
Add"tional in-plant tests of the block valves will be performed at the Virgil C. Summer Nuclear Station, Unit No.1 during the next hot functional testing, currently scheduled for Septanber 1981.
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Robert L. Tedesco, Assistant Di: ector i
8109010612 810819 for Licensing PDR ADOCK 05000395 Division of Licensing G
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- NRC FORM 318 (IO< 80l NRCM O24d OFFICIAL RECORD COPT.
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Mr. T. C. Nichq1s, Jr.
AUOU3T I S 1A1 Vice President & Group Executive Nuclear Operations ---
South Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218 cc: Mr. Henry Cyrus Senior Vice. President South Carolina Public Service Authority 223 North Live Oak Drive Moncks Corner,-South Carolina _.. 29461 J. B. X..otts, Jr., Esq.
Debevoise &_Libennan _.
120017th Street, N. W.
Washington, D. C.
20036 Mr. Mark B. Whitaker, Jr.
Group Manager - Nuclear Engineering & Licensing S',uth Carolina Electric & Gas Company P. O. Box 764 Columbia, South Carolina 29218 Mr. Brett Allen Bursey Route 1, Box 93C Little Mountain, South Carolina 29076 Resident Inspector / Summer NPS c/o U. S. NRC Route 1, Box 64 Jenkinsville, South Carolina 29065 Dr. John Ruoff Post Office Box 96 Jenkinsville, South Carolina 29065 n.
F AUGUST 1 9 1981 j.
BOARD NOTIFICATION DISTRIBUTION: Virgil C. Summer (BN 81-21)
Docket File (50-395)
LB#1 Reading File (Board Notification)
D. Eisenhut/R. Purple NRC PDR Local PDR DL Branch Chiefs W. Kane J. Youngblood M. Rushbrook R. Vollmer T. Murley R. Mattson S. Hanauer B. Snyder R. Hartfield, MPA 0 ELD OIE (3)
H. Denton/E. Case PPAS (H. Thompson)
E. Hughes S. Goldberg/ Young J. Keppler, Region III E. Blackwood, IE Hqrs.
J. Sniezek, IE Hqrs.
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AUGUST 1 0 190f DISTRIBUTION OF BOARD NOTIFICATION (FAILURE OF PORV BLOCK VALVE TO CLOSE DURI'lG TEST CONDUCTED BY EPRI)
Summer ACRS Members Atomic Safety' and Licensing Board Mr. Myer Bender Atomic Safety and Licensing Appeal Dr. Max W. Carbon Board Mr. Jesse C. Ebersole Docketing and Service Section Mr. Harold Etherington Dr. John H. Buck Dr. William Kerr Brett Allen Bursey Dr. Harold W. Lewis George Fischer, Esq.
Dr. J. Carson Mark Herbert Grossmaa, Esq.
Mr. llilliam M. Mathis Dr. Frank F. Hooper Dr. Dade W. Moeller Joseph B. Knotts, Jr., Esq.
Dr. David Okrent Christine l1. Kohl Dr. Milton S. Plesset ltr. Gustave A. Linenberger Mr. Jeremiah J. Ray Randolph R. flahan Dr. Paul G. Shewmon Alan S. Rosen'hal Dr. Chester P. Siess flc. John Ruoff Mr. Davis A. Ward Richard P. Wilson, Esq.
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ATTACHMENT 1 OFFICIAL COPY
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e' SOUTH CAROLINA ELECTRIC a gas COMPANY nose onics son re.
CoLU MBIA. SOUTH CARO LIN A 29218
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Mr. James P. O'Reilly, Director
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. S. Nuclear Regulatory Commission
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Atlanta, Georgia 30303
Subject:
Virgil C. Su=mer Nuclear Station Docket No. 50/395 IE Bulletin 81-02
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Re:
NRC Letter Dated 4/9/81 Nuclear Eng. File 2.8950
Dear Mr. O'Reilly:
IE Bulletin 81-02, " Failure of Cate Type Valves to Close Against Dif ferential Pressure," required identification and action taken or planned for the subject valves at the V. C. Summer Nuclear Station. Table 1 tabulates the required infor-cation for each of these valves.
Briefly, all 3GM38 type valves have received modifications to meet the 2750 psi specification differential. As noted in Table 1, seven of these valves were All moditied, although the valve capability satisfied functional requirements.
4GM88 type valves meet their functional require =ent and do not require modification.
The key to interpreting Table 1, " Potential Safety Consequences," is as follows:
ID No.
Conscouences 1.
(PORV Block Valves) Potential incomplete isolation of pressurizer This is not a safe.y issue oer WCAP 9600.
(FSAR Se'ction e
PORV.
15.3).
2.
Potential cavitation of a centrifugal charging pump or safety injection pump d;e to operation beyond maximum runout flow.
3.
Potential inability to perform post accident coni., inment isolation.
4.
Potential degradation of safety injection flow below values given in SAR.
5.
Potential inability to isolate RCS pressure boundary.
3Gil s
ih 8107240399 10707 PDR ADOCK 000395 O
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Mr. James P. O'Reilly July 7,.981 Page Two Should there be any questions or additional infor=ation desired, please give us a call.
Very truly yours,
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r T. C. Nichols, Jr.
RMC:TCN:glb Attachnent
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V. C, Su=mer w/o att.
G. H. Fischer w/o att.
T. C. Nichols, Jr. v/o att.
H. N. Cyrus J. W. Dixon, Jr.
D. A. Nau=an W. A. Williams, Jr.
R. B. Clary O. S. Bradham A. R. Koon A. A. Smith H. Radin M. N. Browne H. E. Yocom B. A. Bursey J. L. Skolds J. B. Knotts, Jr.
Dr. J. Ruoff R. M. Campbell I&E (Washington)
Docu=ent Manage =ent Branch (55e, 21, LER Only) hTCF File fA.
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ATTACHMENT 2 south CARouNA ELECTRIC a gas COMPANY POST OF MCE BOR 784 Cc' une:A, South CAROLIN A 2 9;'t e T. c. uic ao ts. a a.
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Nucts.e Orteano.s Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 TMI Iter. II.D.1 PORV Block Valves
Dear Mr. Denton:
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At the request of Mr. William Kane, South Carolina Electric and Gas Company provides the following information regarding the PORY block valves for the Virgil C. Sumer Nuclear Station.
As discussed in our letter to Mr. Denton dated 11 arch 25,1981, and in FSAR Section 5.5.13.4, PORY block valves similar to those located at Summer Station s.ere to be tested at Duke Power's Marshall Steam Plant.
Westinghouse type 3GM88 valves, the exact model located at Virgil C.
Summer Nuclear Station, were tested at the Marshall Steam Plant.
Initial test results were unacceptable as the valves failed to clos'e during full flow conditions. As instructed by Westinghouse a design modification was made and subsequent testing proved successful.
The modification invo'ved changing the valve operation from torque to limit control.
As indicated in our response to NRC IaE Bulletin 81-02 dated July 7, 1981, the same modification was made to the Virgil C. Summer Nuclear Station valves.
Additional in plant tests will be performed at Virgil C. Summer Nuclear Station during our next hot functional testing scheduled to start..
in mid September.
These tests will involve cycling the block valve with the associated PORY open at normal operating conditions.
If you require additional information, please let us know.
Yours very truly, T. C. Nichol s, Jr.
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Mr. Harold R. Denton July 28, 1981 Page 2 RBC:TCN:1kb cc:
V. C. Sumer G. H. Fischer H. N. Cyrus T. C. Nichols, Jr.
J. C. Ruoff
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D. A. Nauman W. A. William, Jr.
R. 8. Cl ary
- 0. S. Bradham A. R. Koon M. N. Browne B. A. Bursey J. L. Skolds
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J. B. Knotts, Jr.
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J. B. Cookinham H. E. Yocom
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NPCF File 1
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~ ATTACHMENT 3 SSIN No. 6820 Accession No.
8011040283 IEB 81-02 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 April 9, 1981 IE Bulletin No. 81-02:
FAILURE OF GATE TYPE VALVES TO CLOSE AGAINST DIFFERENTIAL PRESSURE Descr ption of Circumstances:
As a part of its pressurized water reactor (PWR) Safety and Relief valve Testing Program, the Electric Power Research Institute (EPRI) conducted limited testing of a number of valves used on PWRs as power-operated relief valve (PORV) isolation or block valves.
These tests indicate a number of cases in which certain of these valves failed to fully close under conditions that approximated those of their intended service (i.e., saturated steam at approximately 2,400 psi).
The valves that failed to fully close are gate type motor-operated valves that may bc used in various safety-related applications in addition to PORV block valves.
Background on EPRI Testing:
The proposed full-scale qualification testing of PORV block valves, with a completion date of July 1,1982, was first provided to the utilities in a September 5, 1980, draft of NUREG-0737.
The item was formally issued, with Commission approval, in NUREG-0737 on October 31, 1980.
The block valve qualification testing was proposed in NUREG-0737 primarily as an additional means of reducing the number of challenges to the emergency core cooling system and the safety valves during plant operation.
In anticipating a request for PWR block valve testing, EPRI decided to make i
provisions for the installation of block valves betyeen the test steam source and the test PORV in July 1980 at the Marshall test' facility.
The Marshall test facility is a full-flow steam test facility owned by Duke Power Company.
Test PORVs had been carefully selected, with close coordination between EPRI, its consultants and PWR utilities, to assure that PORVs representative of those in service or intended for service would be tested.
However, for the block valves that have been tested concurrently, this selection process was not followed because an NRC block valve test program had not been formulated.
Therefore, seven readily available valves were obtained and tested by EPRI, primarily to obtain some general baseline information on block valve closure capability.
For t e block valves that were tested, EPRI had not established, at least at the time af testing, the population of plants, either operating or under construction, 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 were determined to be a'pplicable for steam testing of PORVs.
=
IEB 81-02 April 9, 1981 Page 2 of 4 These test conditions were selected after review by EPRI, utilities, and PWF.
N555 vendors.
NRC staff also reviewed and concurred with the test conditions.
To date, there has been no similar specific determination by EPRI cr the NRC staff as to the 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.
Westinghouse 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 stampM " WESTINGHOUSE" and jnclude " 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 also determined the threshold differential pressure across the valves above which closure cannot be assured.
These values 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 fu1Ty close.
2.
Borg-Warner Nuclear Valve Division _.(BW-NVD) 3-inch 1500 pound Motor-Operat3J Gate Valves - These valvec can be identified by BW-NVD part nun.6ers 75460, TT9TO, and 79190.
Supplemental testing to determine threshold differential pressures for less severe service has yet to be i
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 l
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 Darling 3-inch 1540 pound Double-Disc Valve - This valve, the first of a series of specially designed valves, has been modified, retested, and demonstrated to close under test conditions.
The remaining valves will be similarily modified during manufacture.
As a result, they are not included in this bulletin.
It must be cautioned that Tables 2 and 3 may not be complete.
For example, the l
staff is aware of one power reactor facility that obtained affected valves from l
another inventory.
For this reason, th s bulletin is applicable to all power reactor facilities with an operating license or construction permit.
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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 number of safety-related
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IEB 82-02 April 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 manufacturer.
For this reason, this bulletin is applicable to the affected valves that are required to close with a diffe ential pressure 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 dif fer-ential pressures of concern include the following:
a.
For the W-EMD manufactured valves, values in excess of the threshold values iii Table 1.
b.
For the BW-NVD valves, any value.
2.
If no affected valves are icentified, report this to be the case and 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 condition I
requiring closure would have on system (s) operability pursuant to the facility technical specifications for continued operation.
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 l
NRC listing the affected valves identified, their service or planned service, the maximum differential pressure at which they would be required to close, the safety consequ2nces 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:
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1.
Ascertain whether any of the affected valves are or will be installed or l
maintained as spares for installation where they are required to close l
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IEB 82-02 April 9, 1981 Page 4 of 4 with a differential pressure across them in safety-related systems or as PORV block valves.
The differential pressures of concern include the following:
a.
For the W-EMD manufactured valves, values in excess of the threshold values in 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 bulletin, submit a report 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 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 accoroance 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 i
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 GA0 under blanket clearance number R0072 that expires November 30, 1983.
Comments on burden and duplication should be directed to Office of Management and Budget, Room 3201, New Executive Office Building, Washington, D.C.
20503.
Attachments:
1.
Table 1 - Identification of W-EMD Manufactured Valves and DiTferential 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 I[ Bulletins t
Attachmont 1 April 9, 1981 IEB 81-02 TABLE 1.
IDENTIFICATION OF W-EMD MANUFACTURED VALVES AND DIFFERENTIAL PRESSURE LIRITS FOR VALVE OPERATION Nominal W-EMD Valve Rodel
" VALVE a***
Size (in.)
Reference IDENT."*
" VALVE I.D."**
(psid) 3 3GM88 03000GM88 3GM58 or 3GM78 or 3GM88 1500 3GM88 03002GM88 3GM58 or 3GM78 or 3GM88 1500 3GM99 03001GM99 3GM58 or 3GM78 or 3GM88 750 4
4GM88 04000GM88 4GM78 or 4GM88 750 4GM88 04002GM88 4GM78 or 4GM88 750 4
4GM87 04000GM87 4GM77 750 4GM87 04002GM87 4GM77 750
_M This number is found on tne yoke-mounted nameplate and occupies the first nine positions of a 24 position number.
It is used in evaluating the functional AP requirements.
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.
Pressere below which valve will close (as shipped).
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 All 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."
number on these indexes or flow diagrams.
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April 9, 1981 IEB 81-02 TABLE 2.
PARTIAL LIST OF PLANTS WITH AFFECTED VALVES MANUFACTURED BY W-EMD
" VALVE IDENT." Number 04000GM88 04002GM88 03000GM88 04000GM87 03002GM88 03001GM99 04002GM87 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
North Anna 1, 2 X
Oconee 1, 2, 3 X
X San Onofre 1 X
Surry 3, 2 X
X X
Zion 1, 2 X
Nonoperating plants (supplied as original scope of supply except as noted):
Beaver Valley 2 X
X l
Braidwood 1, 2 X
X 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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April 9, 1981 IEB 81-02 TABLE 3.
PARTIAL LIST OF PLANTS WITH i
AFFECTED VALVES MANUFACTURED BY BW-NVD Plant NVD-P/N Arkansas Nuclear One, Unit 2 75460 Bellefonte 79190 Palo Verde 77910 i
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IEB 81-02 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 80-25 Operating Problems with 12/19/80 All BWR facilities Target Rock Safety-Relief with an OL & specified Valves at BWRs near term OL & all BWR 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 All power reactor Due to Water Leakage facilities with an Inside Containment OL or CP (October 17, 1980 Indian Point 2 Event) 80-23 Failures of Solenoid 11/14/80 All power reactor Valves Manufactured by facilities 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 l
80-21 Valve yokes supplied by 11/6/80 All light water Malcolm Foundry Company, Inc.
reactor facilities with an OL t r CP OL = Operating License CP = Construction Permit l
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