ML20106J833
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I' ENCLOSURE WO, 1 i
Description of Violations I.
Jersey Central Power & Light Company Medieon Avuaus at Punch Bowl Reed Morriktsen, New Jersey 07908 License No. Brt.16 Docket No. 50-219 Certain activities under your license appear to be in violation of AEC j
requirements.
A.
These apparent vielstions are eensidered to be of Category II l
Severity.
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10 Cyt 20.201(b) spesified that "Each 11eenees shall make or cause to be made such surveys as may be necessary for him to
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comply with the regulations in this part."
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limits shole body exposure to three ren/ quarter.
Contrary to the above requirements, exposure records show that i
one man was exposed in excess of three ran during the April-l June 1973 quarter.
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It was,noted that additional controls over contractor employees j
have been initiated in order to prevent a recurrence.
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2.
Paragraph 4.6.C of the Technical Specification requires that a reactor coolant sample be analyzed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for total radioactive iodine content.
i Contrary to the above requirement, this analysis was not made for approximately 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> between June 9 and June 13, 1973.
l We note that additional controls have been initiated to prevent a recurrence.
5.
This apparent violation is considered to be of Category III severity.
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Paragraph 4.2 of the Teahnical Spesification requires that the oontrol rod drive housing support system be inspected after re-I assembly.
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9604120016 960213 PDR FOIA DEKOK95-258 PDR
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s paragraph 6.5 of the T.S. requires " Records of principal maintenance activities imeluding inspection and repair, of principal items of equipment pertaining to nuclear safety."
centrary to the above requirement, the remeter was restarted tallowing reassembly of the sostrol red drive support systan without any records of the required inspection. We note that to prevent resurrense the drywell sleeure abeek-off sheet (procedure 102.4) has been revised to provide written verification of support systen inspection (uban re-i quired) prior to closure of the drywell.
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M ADisoN AVENUE AT PUNCH BOWL Ro AD e MoRRistoWN, N.J. 07960 e $39 6111
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I September 21, 1973 Mr. A. Giambusso Deputy Director for Reactor Projects Directorate of Licensing
. United States Atotic Energy Commission Washington, D. C. 20545
Dear _Mr. Giambusso:
Subject:
Oyster Creek Station-Docket No. 50-219 ilydraulic Shock and Sway Arrestor Failure The purpose of this letter is to provide infornatica pertaining to additional hydraulic shock and sway arrestor failures at the Oyster Creek stetlon.
A written report as requested in R. O Bulletin Mo. 73-4 uill be submitted upon.cowletion of our inspection prorr.u.
This event is considered to be an' abnormal occurrence as defined in the Technical Specifications, paragraph 1.15.D.
Notification of this event as required by the Technical Specifications, paragraph 6.6.2.a., was made to AEC Region I, Directorate of Regulatory Operations, by telephoac on September 10, 1973, and in writing to Mr. B. Greenman on September 11, 1973, during his visit to the sito.
As comitted to in a letter from Mr. D. A. Ross to Mr. A. Gimbusso dated July 27, 1973, the plant was shut down on September 8,1973 for the purpose of inspecting the hydraulic shock and sway arrestors located on piping systems th m'-bout the dr:~el! rd reneter w ! din,.
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f ailure of 23 out of 66 hydraulic shock and seny arrestors en piping systens in the d2ywell and 19 out of 72 (so far inspected) external to the drywell, some of which are associated with engineered safeguards systems.
All 25 Grinnell shock 1
absolbers on the five recirculation loops in the drywell were inspected and found l
to contain varying amounts of oil. None, however, were without oil and, consequently, are considerud to be operable. They have been refilled as necessary under normal maintenance.
The snubbers were made inoperable due to excessive loss of hydraulic fluid resulting from the failed millable gum polyurethane seals.
,The failed hydraulic shock and sway arrestors are being, or will be,
- replaced using snubbers. rebuilt with seal kits supplied by the Bergen-Paterson -
Pipe Support Company.
Further, all the remaining units in the drywell, al though
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Mr. Giambusso September 21, 1973 noted to be satisfactory during this inspection, are being removed and rebuilt to replace certain critical millabic gum polyurethane seals with nolded poly-urethane seals in the kits noted above. The seals provided to us by Bergen Paterson should provide a longer service life than those previously utilized at Oyster Creek.
The failures were such that they affected both core spray and both emergency condenser systems in the drywell and both containment spray systems external to the drywell. The loss of shock absorber operability results in a reduction in the ability of the associated piping systems to withstand a design bases earthquake. Additionally, failed shock absorber restraints were also dis-covered on the electromatic relief discharge blowdown lines.
Failure of these absorbers would increase the probability of damage to the relief valve dis-r' char e piping during periods of multiple valve actuati.on, s
s The following program is proposed which is intended to provide a permanent modification of our hydraulic shock absorber units to assure their proper long term operation:
We will conduct our next reinspection of our Bergen-Paterson hydraulic y 4, shock absorbers following approximately 6 weeks and no longer than 12 weeks afterfibf the plant has been at operating temperature following this shutdown.
The inspection will include all the items as ioentified in R. O. Bulletin L 70-4 dalud Augu3 t 17, 17,70.
AL the Limu vf the teinspectivn, a prompi Lele-phone report will be made to advise Region I of our findings.
A written report will also be provided as specified in the above mentioned R. O. Bulletin.
Due to the significant radiation exposure associated with the inspection and esmplete rebuild of all our Bergen-Paterson shock absorbers (approximately 20 man-rem to date for this inspection and repair), we believe it most desirable to have the reinspection period coincide with the availability of an ultimate modification, if at all possible.
Our Generation Engineering Department is currently pursuing two equally acceptable long term solutions.
First, General Electric Company has underway a uovetupwent progau wita me snuover anu seal manutaetutet to uetermine onat me permanent satisfactory seal will be oi etnylene propylene material, once this is determined and agreed to be all parties involved, we would purchase enough seal Kits to rebuild all our Ucrgen-Paterson shock absorbers or consider purchasing new Bergen-Paterson units, utilizing the ethylene propylene seal material for installation in the dryuel s.
A second solution currently being pursued is the consideration of re-placing all the Bergen-Paterson shock absorbers presently in the drywell with these of a different manufacturer.
Selection of either alternate will be considered in view of the long term suitability to resolve the hydraulic shock failure problem, tirely avail-ability of material and minimum exposure to station personnel implementing the repair.
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Mr. Giambusso September 21, 1973 We will keep the regional office advised of our progress in arriving at a timely long range solution.
We are enclosing forty copics of this report.
Very truly yours, h,', d.fA?
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Donald A. Ross Manager, Nuclear Generating Stations cs Enclosures cc:
Mr. J. P. O'Reilly, Director Directorate of Regulatory Operations, Region 1 99 e
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,lersey~Centrn) Power /Q, Light Company
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M ADiSON AVENUC AT PUNCH BOWL Ho Ao e MoF4Rl5 TOWN. N.J. 07960 e 539 6111 September 21, 1973 h
Mr. A. Gi ambusso Depaty Director for Recctor Projects Directemte of Licensjur; United States Atotaic Energy ' Conalssion h'ashington, D. C. 20545 SN :. -
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Subject:
Oyster Creek Station Docl:et No. 50-219 4
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The nurnose of this letter is to report a failure of the main staan-1 isolution valve.3038 to meet the acceptable leakage rate criterion as specin cc In Technical Specifi cations 4.5.F.1.D.
This event is considered a violation of the Technical Sp<cifications, paragraph 1.15.E.
This cvent is also considered to be. an abnormal occurrence as defined 2
in the Technical Specifications, paragraph 1.15.E.
Notification of this event as required by tbc Te::hnical Specifications, paragraph 6.6.2.a., was r.ade to AEC Region I, Directorate of Regulatory Operations, by telephone on September 10, 1973 and personally to Mr. E. Greenman on Septecter 10, 1973.
The reacter was shut down on September 8,1973 for the purpose of re-inspecting the Sergen-Paterson shoch absorbers at the Oyster Creek station.
A Icakage rat" test uns conducted on tne nain steam isolation valves in accoraance with proviour contit:.enta to the Atomic Energy Cnnnission.
As a resuh of this testing, which is partially completed, the Icakage rate for NS03B was fout.d to De uppro.umately 400 bbHi based on tue rate or pressure ouA acup uewcen vaivt,3 NS008 and Nmla.
Ice a11cv.<as.le AcaKagc rate lau.1,.as detailed in tim i eu.ni ct:i Specifications, 17., 9.95 SCFH (5'. of Lto [20]). Ti.e other insik valve NS03A Icakage rate was determined to be nondetectabic, <0.1 SCFil. The leakage measule-raents for the outside iso]ntion valycs,elli bc dcterrained once ce i:ava LO.;plotcd th: Curr::.'. l..:p::ti cn an.' ' :i tir c 'T? ?".
Thir. failure is sinitar to one reported ~to your office by my letter i
dated June 5, '1973.
As a result of the failure to achieve an acceptable leakage rate :censur.wmt ct that tire, be JicassenbicJ NS03B, the p.ilot steu.': r re'.;cred c..l - i,. h.'c d '.1 n. u n:
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cc.ntrols.
In nJJition, both the n9 0 seat an.' pilot seat srrf eres rcre rMappN Foj lowing-ren<rembly of t he valve, t he 20 psi nir test indicated no decrable leakage throogh : the t alvo (i.e., <0.1 SCFH).
It was believed at that time that g$
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Mr. Giambusso September 21, 1973 the failure of NS03B to pass the air test was due to the lack of straightness in the original pilot stem.
The original stem, on several previous occasions, was straightened and reinstalled in the valve and ac eptabac aakage test subsequently performed on the valve, llowever, based on our investigation into the recurring leakage problems with this particular valve, it was judged that the stem was re-laxing after operating for a period of time at elevated temperatures, resulting in excessive stem bowing and improper pilot valve seating.
Therefore, two re-placement stems were manufactured by Atwood G Morrill Company to special specifi-cations provided by Jersey Central Power 6 Light Company.
Thc failure which is being reported by this letter reflects the results of the first test subsequent to some operating history on NS03B with the specially manufactured pilot stem.
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A special meeting was held at the Oyster Creek station on September 13, 1973 to review the most recent developments with this particular valve.
The following course of action was agreed upon by Jersey Central Power 6 1,'i ght Company, Atwood 6 Morrill Conaany and two other companics consultine with Jersey Central Pos;cr G Light Company on this problem:
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Prior to Disassemb.linLof NS03B
- 1.. Instrument with dial gaug6 and potentiometer to measure stem stroke at valve closure.
Obtain baseline marks before oper-ating valve.
2.
Instrument cylinder to measure Ap across cylinder.
3.
Perform stroke tests, measure cylinder tn and valve stroke repeatability.
As a part of this, aisc..ieasure stem novement at a junction of cylinder Ap for increments from 6p = 0 to op =
design.
Alt.o check packing friction by loosening cn3 chectir; stem notion and repeatability.
4.
Determine whether stem is 1.istalled such that it is not
" bot toming-out" on top or bottom of operator cylinder.
5.
Check runout of coupling between valve and operator stem.
NOTE:
If recrurenents indicate significant chnnges c'ering the stroke tests, conduct Icakage tests to determine effect on l
valve Icakage.
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Perform complet e diuensional inspection of critical valve parts.
2.
Cylinder examination for obstructions, rust, et c., and con-dition of seala.
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Mr. ' Giambusso September 21, 1973 1
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Refurbish, as required, (Atwood 6 Mcrrill indicated they can provide qualified welders and procedures, if required, for re-stelliting guide surfaces.)
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. Repeat stem stroke repeatability and op measurements, note above, after refurbishment.
C; Long Term Action-It was agreed that a better lapping tool with bearings and in--
ternal supports'is needed.
Atwood G Morrill indicated that such a tool is.being developed by them and is expected to be available in October 1973.
Atwood 6 Morrill will advise Jersey Central Power 6 Light Company of the schedule for delivery of a lapoing -
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Procedures u alnativ.s vuL11..vJ Leve..;111 be peden..;d The luspui.cluas
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under Atwood 6 Morrill's and Jersey Central Power 6 Light Company's l
supervisinn in neenvAnnee with written procedures. These procedures are the responsibility of Jersey Central Power 6 Light Company and i
t will be reviewed by General Electric and Atwood 6 Morri11.
In addition, Atwood 6 Morrill will also furnish a representative to follow this work.
In determining the significants of this valve leakage, the rate of pressure buildup in the reactor was compared to a graph of pressure buildup where at least one valvo in each steam line was leak tight. These plots compared favorably. This implies that one valve in the "B" main steau line (i.e., NSO4L) is leak tight. This was confirmed when pressure buildup between the valves was observed to be approxi-mately the same as the reactor pressure. The redundancy feature will be confirmed
'upon successful completion of the NSO4B leak test.
it is not nossible, at this time. to snecify exactiv vhnt corrective actions.are to be taken to prevent the reoccurrence of this situation.
The course
'of action will be dictated upon con.pletion of the analysis of the extensive di::.cn-
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sional inspection described above.
It is our intention to keep your office in-formed informally through our' Region I compliance inspector; and, following the completion of the program described herein, to forward to your office the written resula, of our insp.;euun, anu tne currecciw activas diaawd iy this inspueolva.
~ h'o are enclosing' forty copies of this report.
Very truly ycers.
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..o Donald A. Ross thmagel, Nuclear Generating Stations j
.cg Enclosures-(cc: ' Mr. J. P. O'Reilly,' Director -
' Dii'ectorate; of Regulatory Operations, = Region 1