ML20117M787
| ML20117M787 | |
| Person / Time | |
|---|---|
| Site: | 05000000, San Onofre |
| Issue date: | 11/07/1983 |
| From: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Robert Lewis, Norelius C, Starostecki R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20114G054 | List:
|
| References | |
| FOIA-84-616 IEIN-83-76, NUDOCS 8505170206 | |
| Download: ML20117M787 (2) | |
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10V 0 71983 MEMORANDUM FOR:
Richard W. Starostecki, Director, DPRP, Region I Richard C. Lewis, Director,_DPRP, Region II Charles E. Norelius, Director, DPRP, Region III James E. Gagliardo, Director, DRRP&EP, Region IV Tom Bishop, Acting Director, DRRP&EP, Region V
.FROM:
Edward L. Jordan, Director Division of Emergency Preparedness
-and Engineering Response.
Office of Inspection and Enforcement l
SUBJECTi IE INFORMATION NOTICE'83-76:
REACTOR TRIP BREAKER MALFUNCTIONS i
.(UNDERVOLTAGE TRIP DEVICES ON GE TYPE AK 2-25 BREAKERS)
The' subject' notice informed licensees of a newly discovered malfunction involving the UV trip device used on the RTBs at San Onofre, Units 2 and 3.
The malfunc-tion may result in the armature of the UV trip device being jammed in an inter-l'~
mediate position rather than in contact with the air gap adjusting screw.
As a result, the RTB could fail to trip on demand within-the specified criteria of
~the UV trip device.
~The notice did not require licensees to take any action; however, it did suggest
~that licensees using the subject breakers mey f_ind it prudent to periodically inspect their breakers to assure that the armatures of the UV trip devices are i
_in contact with the air gap adjusting screw and not in an intermediate position.
.The' notice also. informed licensees that the NRC would consider an RTB to be
. inoperable if the~ armature of a UV trip device is fcund in'an intermediate position.
l' Because of the safety implications of an inoperable RTB, we recommend that l
resident. inspectors encourage affected licensees to examine the UV trip-idevices as suggested in.the notice.
Because of our position regarding operability, we. request that resident inspectors treat any RTB having its UV
-armature.in an' intermediate position as.being inoperable.
Finally, we request-that resident inspectors call us at (301) 951-0550 whenever such a condition is
- found or any additional br,eaker' anomaly is observed.
ortsial Slenell %
E.Q 1mdma h5051gg 840914 ~_
Edward L.. Jordan, Director zDivision of Emergency Preparedness l
BARFIEL84-616 PDRl and Engineering-Response L
Office of Inspection and Enforcement
- Technical
Contact:
I. Villalva (301) 492-9635 cc:
See next page
o Multiple Addressees cc:
R. C. DeYoung, IE J. Taylor, IE S. Schwartz, IE J. Partlow, IE R. Baer, IE R. Vollmer, NRR V. Noonan, NRR D. Eisenhut, NRR F. Miraglia, NRR G. Holahan, NRR 9
Distribution:
IE File DEPER File EAB File
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- IE IVi11alva:mj Rossi orJ 11///83 11h/d83 11/y/83
h Commonwealth Edison g"
, ' One First National Plaza. Chicag1, Illinois
.k ' O 7 Address Riply tm Past Offic3 Box 767 NQ Chicago, Illinois 50690 November 17, 1983
^
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington,_DC 20555
Subject:
LaSalle County Station Unit 1 Request for Expedited Change to NPF-ll Appendix A Technical Specifications Regarding Reactor Feedwater Inboard l
Check Valves Type C Test NRC Docket'No.-50-373-i
Dear Mr. Denton:
,The purpose of this letter is to request the following exigent p
change in. Technical Specifications for LaSalle County Station Unit 1.
CHANGE REQUEST NPF-ll/83-05 l
Exempt the reactor feedwater inboard check valves from Type C test i
requirements until startup following the first refueling outage.
l i
This proposed change is addressed in Attachment A and has l
received onsite and offsite review and approval.
This is required to ollow Unit 1 to startup during the week of November 21, 1983 and it falls l
cithin the Exigent Change category.
I Commonwealth Edison has reviewed this amendment request and has i
determined that no significant hazard consideration exists.
Our review I
'is documented in Attachment B.
Pursuant to 10 CFR 170, this change reflects one example of a L
Class III amendment.
A remittance of $4,000 is, therefore, enclosed.
To the best of my knowledge and belief the statements contained ll herdln;and in the attachment are true and correct.
In some respects these statement's are not based on my personal knowledge but upon infor-mation furnished by other Commonwealth Edison and contractor employees.
Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
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.H. R. Denton November 17, 1983 Commonwealth Edison is notifying the State of Illinois of our request for this amendment by transmittal of a copy of this letter and its attachments to the designated State Official.
If you have any questions concerning this matter, please contact this office.
Enclosed please find three (3) signed originals and forty (40) copies of this letter and the enclosures.
Very truly yours, O N M ehrlo C. W. Schroeder Nuclear Licensing Administrat~or
- Attachments cc:
Dr. A. Bournia NRC Resident Inspector - LSCS G. N. Wright (State of Illinois)-
SUBSCRIBED and SWORN to before me th).s /777V day of ~2/[ 6 _
1983~
~ NotaYy Public 6
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'c ATTACHMENT A LASALLE COUNTY STATION UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST
SUBJECT:
Reactor Feedwater Inboard Check Valves Type C Test RE,FERENCES (a):
FSAR Figure 6.2-32
^
(b):
Technical Specification Page 3/4 6-32
- BACKGROUND LaSalle County Station has a reactor feedwater system primary containment valve arrangement with three (3) valves per feedwater line as
.shown on FSAR Figure 6.2-32 Detail (b) (see attached).
Each line has two (2) check valves and a motor operated gate valve.
This arrangement meets the intent of General Design Criterion 55 on "the other defined basis
= criteria".
During the postulated loss of coolant accident, it is desir-icble to-maintain reactor coolant makeup from all sources of supply.
This design preserves'the reliable coolant makeup to the reactor vessel from
-the normal source when required and prevents inadvertent isolation of the-feedwater lines.
When the plant was originally designed, Commonwealth
. Edison believed that only two isolation valves were required on these lines per prior NRC interpretations of GDC 55 regarding feedwater make-up requirements dominating the isolation requirement.
To meet this design requirement Commonwealth Edison installed the motor operated 1B21-F065A and B feedwater injection valves and the special positive-closing check valves 1821-F032A and B which have~ air
~
operated actuators.
Each motor operated feedwater injection valve was supplied with power from a separate ESS power bus, and can be operated from the control room following a loss of offsite electrical power.
Each of the outboard positive-closing check valves has testability features to conform to 10,CFR 50, Appendix-J criteria.
The feedwater line between the~ inboard.and outboard feedwater check valves as well as the valves themselves were specially designed and-constructed in accord-ence with Standard _ Review Plan 3.6.2-10, and ASME Section 3, so as to preclude the possibility of a credible feedwater line break between the
. check valves-(the "superpipe" criteria).
The inboard feedwater check
-valves 1821-F010A and B were not expected to be tested to Appendix J criteria and were procured with a different leakage tolerance..They were installed only to prevent a significant loss of inventory in the event of
~o feedwater line break.
Subsequently, prior to the issuance of the Unit 1
. operating license,.the. inboard feedwater check valves were required.by NRR to:be tested to 10 CFR 50, Appendix J criteria, as containment isolation valves,-in addition to the other two valves in these lines.
A low pressure seal feature was therefore added to these inboard feedwater
-check valves to meet the Type C test requirement as designated in the Tech Specs..
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. 4 DISCUSSION ~
On November 3, 1983 LaSalle County Station Unit I was shutdown
'for a brief cold shutdown outage to perform various maintenance and surveillance-items.
While testing the reactor feedwater lines per.
1 Technical Specification surveillance requirements 4.6.1.2.d, it was
' determined that the reactor feedwater inboard check-valves 1821-F010A and i
B did not meet' Technical Specification 3.6.1.2.b for primary containment combined leakage rate limit for Type B and C tests (0.6 La).
Commonwealth. Edison currently-anticipates replacing the seal material for the inboard feedwater check valves in accordance with the design-and material specified by the valve manufacturer.
Following the first failure of the valves to pass a local leak rate test which was performed after approximately eighteen months of operation, we observed some. damage to the. original seal material are-concluded was the result of machining the material rather than, which grinding.it to-tolerance.
the pressure equalizing ports in the disk had cut the seal material inWe also conclu multiple locations.
These sharp edges have since been removed.
Because of the difficulty of obtaining molded seal material as was used for'the original seals, the manufacturer supplied new seals 1
chich;were extruded and the ends of the material were vulcanized together l-to_ form the seal.
On November 3, 1983, following approximately another month of operation, the _ inboard feedwater check valves again failed to
{
pts 3 local leak. rate tests.
The inspection of the seals revealed a gap
~in eact seal.on the circumference, one about one-half inch long and the Other aoout one and one-half' inches in length, at the vulcanized points l-
'of1the seals.
The seal material in the "B" valve also appeared to be i
brittle with multiple minute cracks.
We are also investigating a possible slignment problem that-may have prevented the disk from closing squarely against the seat which.could also contribute to type "C" test failure.
The valve manufacturer is now supplying Commonwealth Edison with new' molded seals as in the, original design.
We believe this will resolve the new and different failure modes which were experienced with the extruded / vulcanized-seals.
A representative of the manufacturer is also
(
on-site to evaluate.the effects of the alignment tolerances.-
With these repairs, we believe that the seal design and material L
in the inboard feedwater check valves will be essentially identical to the seals which have been successfully used and tested at other si,tes for L
pariods of several years.
ife also anticipate that a successful local j-loak rate test will be passed followin on the inboard feedwater check valves.g the repairs currently in progress V
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.- ' However,.because of two successive failures of the valves to
. pass the local leak rate test criteria following operations, we feel that further relief from the Appendix J criteria should be sought for these valves until' solutions to the existing problems can be resolved and further-repairs and/or modifications can be performed during the-first refueling outage.
CONCLUSION
. Because the.feedwater lines will still'have two isolation valves in each line which meet the requirements of Appendix J to 10 CFR 50, because the outboard motor operated isolation valve is supplied with power from an.ESS bus, and because the feedwater line between the two check. valves is designed and constructed so as to preclude a credible
,line break, it is believed that no unreviewed safety hazard. exists and that compliance with GDC 55 is not compromised.
For LaSalle Unit 1, therefore, Commonwealth Edison proposes to change the Appendix A Technical Specifications to License NPF-ll as indicated on the attached marked up page 3/4 6-32 until startup after the first refueling outage.
This amendment will exempt the Feedwater Line Inboard Check Valves from Appendix J -Type C testing until startup following the first refueling outage.
The leakage requirements of Appendix J have been met by currently valid Type.C tests performed within the past 3 months on two valves in each feedwater line:
Feedwater Outboard Testable Check Valves
'(Item #19,'1B21-F032A and 8, on Tech Spec Page 3/4 6-27) and Feedwater Outboard Remote Manual Isolation Valve (Item #2, 1821-F065A and B; on 1 Tech Spec Page 3/4 6-32), and Reactor _ Water Cleanup Return Valve (Item
- 2,'lG33-F040,oon Tech-Spec Page 3/4 6-32).
In addition the remote motor operated valves 1821-F065A and B and RWCU valve 1G33-F040 will be admini-stratively controlled as stated in the LaSalle FSAR Section 6.2.4.2.1:.
"The valve (s) is remote manually closed-from the main control room to provide'long term leakage-protection upon operator determination that continued makeup from the feedwater system is unavailable or unnecessary (after LOCA)."
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AMENDMENT 47 OCTOBER 1979
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RPV Containment A0 A0 J
B C
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DETAIL r,
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FEEDLJ4 TEM LI NK
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NOTE: TC DESIGNATES TEST CONNECTION.
LA SALLE COUNTY STATION FIN AL S AFETY AN ALYSIS REPORT C
FIGURE 6.2-32 CONTAINMENT VALVE ARRANGEMENTS (SHEET 1 of 10) e m
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trd', b '!,1 1 w:s TABLE 3.6.3-l'(Continued)
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PRIMARY CONTAINMENT ISOLATION VALVES y,
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- VALVE FUNCTION AND NUMBER -
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Other 'Is'olation Valves
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, Residual Heat Removal / Low Pressure Coolant Injection System I
1E12-F042A,' B, C,,
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1E12-F017A,*'BJ*f3)
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'1E12-F027A;'BId) h1E12-F024pj)B IE12-F021
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I1E12-F064A'48 II)
'.1E12-F011A,'Bqg II)
IE12-F088A; B, C II) 1E12-F025pj')B,C l
'IE12-F005(3)'
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1E12-F055A,'B(g3),,:.,,,,
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1E12-F311A, B s
IE12-F041A,B(kS(k)-
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IE12-F050A, B u'huif.s to:trmunu j a 2., ia. certi.
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ATTACHMENT B a
Significant Hazards Consideration Commonwealth Edison has evaluated the proposed Technical.
Specification Amendment and determined that..it does not represent a
- significant hazards consideration.
Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92,. operation of.LaSalle County ~ Station Unit 1 in accordance with the proposed amendment eill not:
1)' - Involve a significant. increase in the probability or consequences of an accident previously evaluated because of two (2) other valves in the line:will be type C tested per the Technical Specifications, FSAR and Appendix J to 10 CFR 50 criteria and Commonwealth Edison believes this change is an acceptable alternative to General Design Criteria 55.-
- 2) ~ Create the possibility of a new or different kind of accident from
~
s anyLaccident previously evaluated because:
a)
The containment-leakage criteria will still be met even with single 1 failure-since two valves in the line will meet Appendix J
- criteria.
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b)
'The feedwater line break accident will not be affected by this change because the inboard check valve will still be able to prevent. gross inventory-loss if a feedwater line break were to occur.
1 3)
Involve a.significant reduction in the margin of safety because the leakage criteria for the primary containment as a whole will still-meet the leakage margins sus required by Appendix J to 10 CFR 50.
. Commonwealth Edison believes that-two valves in each line which meet-Appendix J criteria is an acceptable alternative to General Design Criteria 55 on the "other defined basis criteria."
Based on the preceding discussion, it is concluded that the iL proposed change clearly-falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated cccidents will not be increased and the margin of safety will not be
. decreased. 'Therefore, based on the guidance provided in the Federal
- Register:and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration.
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.o 1, : o NOV 2 e 23 DocketNo.50-10'lh f
Docket No.-50-237e2 -
DocketNo.50-249'lh Commonwealth Edison Company ATTN:
Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:
Enclosed you.will find a revised first page for the Notice of Violation
. submitted to you with our letter dated October 13, 1983. This revision corrects an error in the inspection period referenced in the Notice.
hewillbehappytodiscussanyquestionsyouhaveregardingthismatter.
Sincerely, "Qriginsi signed,by U. D. Shafcr" W. D. Shafer, Chief Projects Branch 2
Enclosure:
Revised Page for Notice of Violation Issued October 13, 1983 cc w/ enc 1:
D. L. Farrar. Director of Nuclear Licensing D. J. Scott, Station Superintendent-DMB/ Document Control Desk (RIDS)
Resident Inspector, RIII Phyllis Dunton, Attorney General's Office, Enviro.nmental Control Division
.i F311.79 @'t 11_-
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1/10/83
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Appendix Y
NOTICE OF VIOLATION Commonwealth Edison Company Docket No. 50-237 As a result of the inspection conducted on July 20 through September 19, 1983, and in accordance with the NRC Enforcement Policy, 47 FR 9987 (March 9, 1982),
the following violations were identified:
1.
10 CFR 50, Appendix B, Criterion XVI, states in part, " Measures shall be established to assure that conditions adverse to quality,...are promptly identified and corrected.
In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition."
Commonwealth Edison Company Topical Report CE-1A " Quality Assurance Program for Nuclear Generating Stations," Section 16, states that corrective actions are verified for satisfactory completion to preclude repetition.
Contrary to the above, adequate corrective action to preclude repetition was not taken following a July 19, 1983, incident where a perforation was found in a' corrugated expansion bellows on one of the torus-to-drywell vacuum breaker lines causing a breach of primary containment. This resulted in three more perforations on two other expansion bellows dis-covered on August 11, 1983.
This is a Severity Level IV violation (Supplement I).
t-2.
Technical Specification 6.2.A.7, " Plant Operating Procedures," requires that detailed written procedures including applicable checkoff lists covering surveillance and testing requirements shall be prepared, approved and followed.
i.
A note in Surveillance Procedure DOS-1600-1, "LPCI System Valve f'
Operability Test," states, " Drain water between 1501-27 and 1501-28 L
before opening 1501-28 valves to prevent possibility of stagnant water plugging drywell spray nozzles. Leave control switches in normal L
operating position. Valve 1501-28 may be run after water is drained."
Surveillance Procedure' DOS-1500-1, " Quarterly Valve Timing," also tests the 1501-27 and 1501-28 valves.
I Contrary to the above:
i Surveillance Procedure DOS-1600-1 was not sufficiently detailed in a.
that it did not include the precautionary note concerning draining the water between valves 1501-27 and 1501-28 before manipulation!of I
those valves.
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