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{{#Wiki_filter:- s afCIDetRE No.1 ) BaseE1E11 p l S C B A 4 ti_on f Jersey Central Power sad Light Onupeny Madison Avenue at hasch Bowl Road '1 Morristeist, New Jersey 07940 Boekat No. 50-319 License No. Est-16 One estivity under your 11eense sypeers to be in violation of Asc seguire-monts as indicated below. This apparent violation is seasidered to be of Category 11 esverity. Paragraph 6.6.2 of the Technical specification requires that you notify the Director of Regional Regulatory Operations Office in the event of an abnormal occurrence and that this notification be made by telephone and telegraph within 24 hours of your recognition of the unusual occurrence. It also requires thst you submit a writ-ten report of the occurrence to the Director of Licensing within 10 days. An abnormal occurrence is defined, in Section 1.15 of the Technical specification, as a failure of one or more couponents of an engineered safety feature or plant protection systen that causes or threatenes to cause the feature or system to be incapable of per-foitiing its intended function. 1 1 Contrary to this requirement, you failed to notify the Director of the Regional Regulatory Operations Office, or report to the Director of Licensing, within the prescribed time limit s, that 88 of 132 shock l suppressors had been found defective between April 15 and Jut.e 5,1973. l 22, 1973 you failed to make timely notification and to Again on July l submit a timely report when you discoexed that 8 of the reconditioned l shock suppressors had again been found to be defective. We note that tha.e matters were ultimately reported to the t>irectorate of Licens-ing in your letter dated August 6, 1973. I e i i i i t 9604150021 960213 PDR FOIA i DEKOK95-258 PDR
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k' p ..d. c ..t _ \\ . Jersey Central'P' owe 8& Light Company ,/ i n%& M AolSoN AVENUE AT PUNCH BOWL Ro Ao e MoRRibToWN, N.J. 07960 e s39 6111 ' October 12, 1973 j I M v 3 Mr. A. Giambusso C D Deputy Director for Reactor Projects Directorate of 1,1 censing i y. / .7 % g '.1 United States Atomic Energy Commission
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Dear Mr. Giambusso:
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Subject:
Oyster Creek Station 4 g/ Docket No. 50-219 Main Steam Isolation Valve Failure The purpose of this letter is to report a violation of Technical Specifications, paragraph 4.S.F.1.D., failure of main steam isolation valves NSO4A and NS04B to meet acceptable leakage rate requirements. This event is -also considered to be an abnormal occurrence as defined in the 'lechnical Speci- .2 1.15.E. Notification of this failure, as required by the l fications, parar Technical Specf ations, was made to AEC Region I, Directorate of Regulatory Operatit,ns, on Friday, September 28, 1973 and by telecopier on that same day. Following completion of extensive maintenance and repair work on main stcam isolation valve NS03B, it was possible to conduct a leakage test on the two main steam isolation valves (NSO4A and NSO4B) outside the drywell. As a result of the ensueing test, main steam isolation valve NSO4B leakage rate was determined to be 15.2 SCFH and isolation valve NSO4A laakage rate measured 96 SCFH. It was necessary to operate both valves in order to provide adequate ventilation of the reactor vessel while performing maintenance work on NS03B. Thus, the outside isolation valves were operated after the plant was shut down and were not tested in the "as found" condition, as is normally the case. The air flow path established by utilizing the main steam lines was successful in minimizing the radiogas concentrations in the drywell; thereby, providing maximum ~- radiological protection for maintenance people while repairing NS03B. Investigation into the cause of leakag'c through the two outside isolation valves resulted in identifying the valve stem packing as the leakage path. Re-placement of the packing and subsequent retesting of the valves indicated essentially all the Icakage associated with NSO4A and NSO4B was through the stem packing region. As an additional precautionary measure, the two inside main steam isola-tion valves were also, repacked. ?G34 / 5 l U NqDm > ~. --- hp. ,f ~ O J~
i w. a s. n. t i Mr. Giambusso October 12, 1973 In our. letter dated September 21,-1973, we indicated that the safety significance of the failure of NS038 depended on the condition of the outside valve in the "B" steam line, i.e., NS04B. With the 'fa' lure ot'.sSO4B to pass a leak rate test, neither valve in the "B" steam line was capable of satisfying the Technical Specifications leakage rate limit of 9.95 SCFH. 1 It should be recognized the leakage through the packing of NSO4B was i equal to S.70 of the total allowable primary containment leakage; whereas, the allowable Technical Specifications leakage from any one penetration or isolation valve is 5% of this total allowable leakage from the primary containment. There-fore, in the event of. a LOCA, release of. fission products from the primary contain-went would not be greater than the release discussed in Table I.5-2 of Amendment 65 and Section 3.3 of Amendemnt 68 in the FDSAR. It should be noted that the leakage through the packing,of NSO4B would, for-the most part, be. drawn -into the reactor building ventilation system and re- . leased through the plant stack. The failure of NSO4A represents a failure of one of two redundant valves in the main steam line "A". Leakage through the packing of NSO4B would also be into the reactor building ventilation system and would still be a controlled release under hypothetical accident conditions. It should be noted that this is the first time significant stem packing leakage existed in the ouIer main steam isolation valves. Based on past experience with the main steam isolation valves, a failure of this nature has not been previously experienced. Considering this, we intend to investigate a preventative maintenance program whereby a schedule of complete main steam isolation valve inspection can be accomplished. A sct frequency will be determined for this insp'ection in order that all four (4) main steam isolation valves be checked within a reasonable time period. This program should preclude future failures of this kind by identifying problems prior to their reaching a point where degradation of valve integrity occurs. Enclosed are forty (40) copies cf this report. Very truly yours, - (Lh 6%LL ^ Donald A. Ross Manager, Nucicar Generating Stations s DAR:cs Enclosures L .cc: Mr. J. P. O'Reilly, Director Directorate of Regulatory Operations, Region I f.: 1 i ' i
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i-i JEREWY OFNTRN POWE A LIGHT QQMPANY, + g' NUCLEAR DENER ATING STATION DYSTER CREEK ~ --] e.o. sox 368 e FoR#.Eo RIVER e NEW JERSEY e 00731 l nnonE res e r.sw;st y, j October 11, 1975 i i } Mr. Jane.4 P. O'Reilly Dltretorate of Regulatory Operations l Region 1 1 . U. S. Atomic Encrgy Commission. 631 Park Avenue ~~-"
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. --. King of P russi* r Pennsylvania -. 19406 -' *"* -
Dear Mr. O'Reilly:
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Subject:
Oyster Crvek Nuclear Genersting Station i Docket No. 50-219 Preliminary Abnormal Occurrence I;a' port No. 73-26. Per a telephona conversation between.J. L.. Sullivan, Jr., and. D. Caphtyn on October 30,.1973, we arer reporting the attached event as an abnorral occurrence,. although' it~ is, not: clear that it is reportabic. t ? Technical Specification.4;1, Table A.I.1, Note.2,'. states, "At. least daily during reactor. >ower:Uperation, the reactor. neutron flux peaking factor r. hall be estixwted and the flow-referenced APRM scrm and rod Slock settingr shG.be adjusted,'if necessary,.as specified.in Section 2.3, Specifiaw f ons (1). (a) and (2)-(a)." This. estimate was, in fact, performed As specifiedisnd ~ corrections were made as required. - Very truly yours, / h T.Fot. h ..Carro11', Jr. Station Superintendent JIC/pd cc: A. Giambusso 1 y m z~ =
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y.. 2.. -. y p.. 2 g.% .apy .4 = ..10,1973, at 4:30 p.m., and - Operations by telephone, on October by telecepter on October 11 1973;.at31:15 p.m.- SITUAT10m On October 6,1973, at 2:00'pm...the~ reactor startup to full; powe'r had benn halted dWTo~if 1Eck Of71MU$CTV1cc conden?.hte de ' 1 nincratizers. The. core therwl output at this time;was approxi-8 rately 567 M1ft:and ths. recirculation flow' rate.was 30X10.lba/hh At~ this time the Maximun Total Peaking Factor -(PF) was estimated to be 4.54 and.the Average. Power Range Monitors. (APFWs) were set conservatively,such that 100Lon. the APRWs corresponded to - 1200 MWt.. This is equivalent to reducing the neutron. flux scram s by the pount 3.01/PFnas.specifiedeln Tcchnical Specification - 2.3.1.n, with some added margin. The sICO' /1200. Mft setting. allows for a noutron flux peating uptio a value.of 4.84. J ]
I Report H h 7F26 s " 2- ' 0ctobst 6i.'1973 s ~ SITUATION - Continued r At 5:30 p.m., after a heat"balsnce / calculation, the setting of ll_ W L w 21""iT S M ;2 3 ..gc.wan.xtrasse&.s,i1':TQA*nwWPBE= K15iwm 2+ M yswglc23 \\ .waar'~T T-13 7 '. - t hat.1100 t.: ofr that. APRM's===se r.icr.2 f! ; ' og ?tthhb 4 s.udvirg.qly stf.sh% 41tgt lk,% % -tig., 174n ~ i -- ---corresponded to 1400 MWt which accounts for peaking factor $ 'of _ _ =. -only 4.15. 'Ihus, the limiting safety system setting for the ' . APFJ4 Neutron Flux Scram and rod' block were set less conserva-tively'than specified in the..Technica1' Specification 2.3.1.s and 2.3.2.a. CAUSE: An investigationTis yet to-be conducted to determine the exact
- came of this occurrence. Ihowever, at this time, it is' believed i
l .ithat.it was caused by a commication probler:;. t. 9EftDI AL. _ ACTION: ' At:10:00 a.m. on October '7(1973, the reactor. neutron flux peak. ing factor wns estimated as: required in Technical Specification .4.1, Table.4.1.1, Note 2, and(found to bed,71. Tho' APP.Ll's were then correctly adjusted to the conservative'100%/1200 W t setting. SMETY SIGNIFICANG: Based on the Neutzun Flux Peaking ' Factor of.4.71, as estimated. ~ at the tino of the correction, the safety' limit can be shown to ' be at 1226 MMt for the recirculation ilow rate of'30X106 lbm/hr. i Using the 100V1400 mt setting of the APRH's, the reactor at this condition would have scranned at 1200 Wt, if required. -Thus7 the sofety-limitwouldsethevel.a.. acceded - 4 a 2.iG-i.~,-2Wh hh.* kE4!2y $dmana-- A ' ' * ' ^ ^ ^ l d '1 a w s,.. m,' h EPRidtEChyC:; n.w...- =,", + n.wE7 E+==41ie: 8I M =E ...e -==.-+ ~ ~~+~ me .e-mu s o i.,.A.cm.c%n=rJ, s m:nmwtirdbr,rz Jma-. w mu.,rs *Jir#/.r-~-- ~~" *""~" "" *"- ~' **~:
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