ML20107C359
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To:
James P. O'Reilly Directorate of Regulatory Operations Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Fromi Jersey Central Power G Light Company Oyst,er Creek Huclear Generating Station Docket #50-219 Forked River, New Jersey 06731 l
Subject:
Abnormal Occurrence Report No. 50-219/74/ 20 The following is a preliminary report being subndtted in compliance with the Technical specifiestions paragraph b,6,2, Prolininary Approval:
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Date cc Hr. A. Glaubusso l,
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htltial Tolophone
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- Date cf Rcport Datu;
, 3/11/ 74 Occurronect 3/10/74 initial Written Timo of Report Date 3/11/74 Occurrence:
1259 OYSTER CREEK SIR: LEAR GENEltATING STMt0M FORKED HTVi!R, NEW. JERSEY 08731 Abnormal Occurrence Report No. $0-219/74/ 20 IDENTIFICATION Violation of the Technicni Specifications, paragraph 4.54.1.d, OF OCCURRENCE:
failure of Main Stoan Isolation Valvos NSO4A and NSO4B to meet the allowablo leakago requirements.
'This event is considered to be nn nbnormal occurrence as dra fined in the Technical Speelfications, paragraph 1,15nJ__Fm CONDITIONS PRIOR TO OCCURRENCE Steady State Power Routinc Shutdown
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llot Standby Operation Cold Shutdown Load Changes During Refueling Shutdown Routino Power Operation.
Routine Startup 1 Other (Specify)
Operation See below The plant was shutdown with the reactor coolant at <212*P, with the reactor undo in REFUEL.
DESCRIPTION The 65tV's were tested in the "ns found" condition to the OF OCCURRENCE:
extent that the valves weren't cycled before the test. The volves woro not, howcyor, closed under pressure.
1159 - Leak rate tests on N31V's NSO4A and NSO4B hegan c1239 tmak rate ' tests'on SGIV's NSO4A and NSO4B ended cf 3 o n w n q c t__3 p P,
Report CSo. M-8DB/73F
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14Akage rates of HS1Y's NSO4A and 'NSO414 woro 64.7 SCIH and 12.2 SCFil, corrected to 20 psi. The maximum alicwable leakage rate in 9.945 SCFil, as required by the Technical Specifications,
- paragraph 4,5,F,1, d, APPARENT CAUSp.
De81gn Procedure OF OCCURRENCE:
~ houfacture Unusual Service Condition
~ Installation /
s inc. Environwntal Const ruetion X "Composeiit Fal10ro
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Operat.or Other (Specify)
After checking the test W.ne*1y and the components of the MS1V's, it was determined that the Icwor packing ring around the valvo shitft was the cause of the excessive leak rate. The leakage was out of the loskoff line betwcon the upper and lower sets of packing. A cause of repetitive leek 3Re usy be thst, when the valve is repacked in place, the packing ring ir. cut i
and then installed around the shaft. This packing is designed' for installation by sliding the packing rings down the shaft without cutting. The valve op trator must he dinconnected for this method of installation, AN/M$1$ OF 1hc sitfety significance of the failu're of NSO4A and NSO4B to ocammpCn:
pitss the loakage rate test was a loss of redundancy in an 1
engineered safety feature designed to minimite the release of fission products under design bases accident conditions, it should be noted that any lenkage through the lower set of packing would be into the reactor hullding equipment drain tank
Weport no. ov-sui nm t.
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and would be released through the plant stuck via the standby gas treatment system.
It should also be noted that the inside MS1V's are, reported to havo no detectabic Icakago and,.therefore, there would be no leakoge ou'. of the primary containsment, CDRRECTIVE The 161V valve shaft packing leakoff valve (between upper and ACTION:
lower sets of packing) were closed and the FEly'.4 were retened sucs;er.s tully, 1hese valves will remain in the closed position until the 1974 refueling outage, at which time the volve shafts will be repacked. NS03A and NS03B woro retented to inr. ore that they had an acceptabic leak rate,.since.the 1,est assumes that the velves, NSO4A and NSO4n, have negligible leakage.
'the retests of HS03A and NS03B indicated no detectable leakage.
FAILURE DATA:
'1hc vs1ve stem packing on NSO4A failed en Septenhor 27, 1973
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and ngain un Janunry 16, 1974 Each time, the valvo was re-packed and sibsequent'
)assed its Icak rate test. 1he valve 8 tem packing on NSO4B. ailed on Septenbor 27, 1975 and was sub-sequently repacked, Prepared by:
W Dator 3/11/74
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To t' James P. O'Reilly Directorate of Regulatory Operations Region I 631 Park Avenue v
King of Prussia, Pennsylvani A 19406 i
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From:
Jersey Central Power 61.ight coinpany I
Oyster Creek Kuclear GeneratinP, Station Docket #50-219 Forked River, Rex Jersey 08731
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Subject:
Abnormal Occurrence Iteport No. 50-219/74/ 1_9 _
The following is a preliminary report being subaltted in compliance with the Technical Specifications i
paragraph 6.6.2, Preliminary Approval:
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3l11/74 h(, Carroll, Jr. V ~
Oate cc: Mr. A. Giebu.<,.so k
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unnrarnstnsam unus as Htport Dato3 3/11/'4 Occurrence:
0530
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QYSTER CREEK XIJCI. EAR GENEP.ATING STAT!GX
' FORKED RIVElt, NEW JERSEY 08731
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Abnormal Occurrence l
'. Report No. 50-219/74/19-IDENTIFICATjQN VAclation of the Technical Specifications, paragraph
_3. 5. A.1, OF 00CURRENCE:
failure to maintain primary containnent integrity with reactor water temperature above 2I2*P-and. fuel in the reactor vessel,.
'this event is considered to be an abnormal occurrence as de-
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fined in the Technica1' Specifications, paragraph 1.15B CONDITIONS PRIOR TO. OCCURRINCE:
Study state Power Routine Shutdown llot Standby
~~ Operation Cold shutdown Load Changes During
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Refueling Shutdown
~ltoutine Power Operation l<outine Startup
_ Other (Specify)
Operation 0
The plimt was shutdown with reactor coolant <2121.
I DESCRIPTION At 1000 on March 8,1974, an orderly shutdown of the plant j
OF OCCURRF.NG:
commenced to perform maintonenco on six of the fourteen torus to drywell vacuinn breaker valves. Although thirteen of the valves were considered to be operable at this time, the plant was shutdown in order to effect more permanent repairs on the valves (see Abnormal Occurrence Report No. 74-16, dated I
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koport. No,; WO-2)D/70/19 N^
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March Sj - 1974).
In accordance with the requirements of para--
graph. $ 3.5.A.I'of the Technical Specification % maintenance on these valves ~ did not begin until reactor coolant. tempersture
.was below 212*F.
At approximately 0520 on March 9,1974, a radiation protect. ion technician reported to the Shift Forenen that water vapor i
i appeared to be issuing' from a special manometer which liad been installed for monitoring of the pressure difference between the drywell and reactor ' vessel. 'the Shift Foreman's investi-gation revealed that the reactor sido of the manometer was hot.
The recirculation loop temperature recorder, which wa.e being l
used to monitor reactor water temperature, was inmodiately re-l checked. This recorder indicated a tepperature of 1604 llowever, the indicatinMjumped to approximately 250F when the 5
recorder was bumped. At this time (0531 on March 9,1974) shutdown cooling sy8 tem flow was increased to decrease the reactor water temperature.
Reactor water temperature ww: re-duced to Icn than 2)fF within tipproximately 30 minutes.
Within approximately 130 minutes, e reactor water temperature
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of 1(G*F was established and maintained.
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i APPARMT CAUSE X
Design
_ Procedurc QF OCCtlRRDE.
Manufacture Unusual Service Condition
' ~ ~ InstallatJon/
Inc, Environmental Construction Component Fal. lure 1 Operator Other-(~5peci fy)
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HopoEU CA% Mi-2J9//4/19 Page 3 1
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This abnormn) occurrence is attributed to equipment malfunction and operator error. The recirculation loop recorder did stick and givo a false indication of reactor coolant temperature, liowever, the control room operator failed to react properly to indications that reactor water tehrporature was increasing.
Specifically, a review of the chnrt pnper frod)the recorder monitoring shutdown cooling system temperatures showed that the "C" loop best exchanger inlet togorature was increasing at n rate of about;10'P/hr during}the three hour period prior to 0230 on March 9,1973 At this time, the control rcom operator secured flow in this loop and thereby contrJbuted to the risc in reactor water temperature.
/WALYSIS OF 1he primary containment ' system provides it harrier against OCCimFGNCTi:
wicontrolled relesie of fW;sion products to the environs in the event of a break in the reactor coolant systems.
hhenever l
the reactor coolant water temperature in above 212'F, failuro of the reactor coolant systein could cause rapid expulsion of the coolant from the reactor with an associated pressure rise in the priwary containment.
Primary containment is required, therefore, to contain the thernal energy of the expelled coolant and fission products which would be released froin any fuc1 failurus resulting from the accident.
The safety significance of this event is that primary contain-i ment integrity was not maintained during the period that the reactor coolant tempervture was in excess of 212*P due to the
R CJ% o @ Tecurr:nce
,,.- p port No. 50-219/74/10 PaBe 4 3
saaintenance being perforsed on the vacuun breaker valves. At the condition that existed, the safety signifiennce is consid-cred' minimal.
CORPJM21W The follcuing remedini actions will be taken prior to the.P0 llc AC1'10N:
t evaluation to precludo a recurrence of.this type event:
1.
Control room operatort. Will be instructed to "jogu any recorder that is producihg a suspiciously straight trace.
1his will be accomplished by moinenturily turning the re-corder off and then on again.
2..
Involved personnel will be reminded to utilize all avail-able indicators when vonitoring critical parameters such as reactor water tenperature.
3, The shutdown log will be reviewed and n,ndified to require the recording of additicnal system togeratures which vre relat.ed to the reactor coolant teneraturo.
FAILURE DATA:
Basie recorder data are as follows:
Manufacturer - General Electric Type - GF./MAC 531 Span - 4 inches l
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Date:
3/11/7_4 l
Prepared by
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To:
Jamo$ P. O'Rel31y Directorate of Regulatory Operations Region I j
631 Park Avenue j
King of Prussia, Pennsylvania
'.19406
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From:
Jersey Central PoWor Q Light company Oyster Creek Nnclear Generating Station Docket F50-719 Forked River, hw Jersey 08731 l
Subject 2 Abnormal occurrence Itoport No. 50 219/74/ 18 h following is a prolininary report being submitted in corepliance with the Tecimical Specificationr.
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paragraph 6.6,2, Preliminary Approval:
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[7. T.' Carroll, Jr. 7 Date cc: Mr. A. Gianbusso e
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7 Ur$Y1/h (:ft!!!n: :;Uhl! AIL O!!s'fdtlitiRG f'TNrTO;c F0ntE9 Jili'Elt, NDi JLit%Y 00731 Abnornal Occurrence nepor'; lin. %-2.19/N] 18
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3Dr.UI)) CAT 30.\\
Viointion of the Technical Specifications, phraf:nph,)1]/A_,
O f O f.t tQ ni!.st L:
Failuro of Borgen-Paterson tiydraylic Shock and Sway Arrestors in the drywell.
%11, event i!. cm.i& red to be im nhnornal ncturrence as dn-fined in the Technievi'Specificationu, parvipaph,:,Isn,__,
CONDD10.tD PiEX 70 (KCN.imi:Cl!:
fa.o.9dy f; trite Power Routine f.;hutdos:n
~~" IM St,n dby Opcynti sm "T Cold Shu dnwn
, Loud (:henyt, Da ring ikfut Hnt: Shut dmen llout.Ine Poier 0;*nt), on
(({ houtini Stnytup Other (Specify)
Oporati on The plant, was shutdown with reactor coolant at <222 F, 6
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IESCidl'D Ok An inspection of the drywell snubbers, Borgen Patorson typo 0.F O2(;m:hadv:
HSSA-10, located three'inoperabic units and four which were leaking. They are os follows:
'M3501 #2 - A 18olation Condenser Failed 487574
- B Isolation Condenser Loeking 487502
- B Isolation Condenser Leaking 487495
.- Cleanup System leaking 487573
- Shutdown Cooling Psiled 4874E9
- North Electromatic Relief - Leaking 487446
- South Electromatic Relief - Failed All of the above units, with the exception of 487489, were rebuilt in January 1974' with I?P seals.
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d.n Pro' Jurn APPA10.h'f CMl?.P.
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~~4?dluf fM tuve
[ Min.us).Serviet: Conditaon
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Inc, lhwir' nnental o
(1sy;fs1.ructi Du CtMt'p0D ctil, li j lu ru
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_ tyierstra Other (Specify)
We cause of snubber inoperability was a loss of the hydraulle
- fluid, An investigation is being initiated to determine why the fluid was expelled, 1
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ANM,Y!;1! Of he safety sig'nificance of this occurrence was a partial loss OCm iHG:
of the seismic restraining ability for the affected systems, Hed the plant suffered a design bimc8 corthritnke, the probability that these systems would have iuffered structural dtmage was increased.
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The failed units were replaced with identical snubbers which l.O Stut.
.were rebuilt with othylene propylone seals.
J 1 A).!M: liATA:
M;uiuracturer: Bergen-l'aterson Type: llSSA-10 J euende 8"MP8'88 0 g W Weemmes 3
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