ML20151C171

AO 73-29:on 731103,while Inspecting Hydraulic Shock & Sway Snubbers,Accumulators on Steam Line to Isolation Condensers a & B Found Devoid of Fluid.Cause to Be Determined Upon Insp.Condenser a Refilled & Condenser B Replaced

ML20151C171
Person / Time
Site:	Oyster Creek
Issue date:	11/05/1973
From:	Reeves D JERSEY CENTRAL POWER & LIGHT CO.
To:
References
AO-73-29, NUDOCS 8103040859
Download: ML20151C171 (2)

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)) VEIN 10NT YANKEE NUCLEAR POW Mn COlmOR3 ATFON

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SEVCNTY SCVEN OROVC STRCCT RUTLAND, VI<itM'O NT 0 5'701 REPLY 10:

VYV-3071 p, o, gox ,37 VERNON, VERMONT ol.751 November 14, 1973 Director Directorate of Licensing United States Atomic Energy Commission Washington, D.C. 20545

REFERENCE:

Operating License DPR-28

• Docket No. 50-271 Abnorcal Occurrence N6. A0-73-31 Gentlemen:

As defined in Section 6.7.D.1 of the Technical Specifications for the Vermont Yankee Nuclear Pouer Station, we are rcporting the following Abnormal Occurrence as A0-73-31. .-

On Novenber 7, I973, at 2101, while the plant was in a shut down condition and while the required Control Rod Friction testing uns being perforned on control rod 26-23, a react or scram occurred initiated by a high-high flux signal from the Intermediat e Ranac Neutron "onitoring System.

An inmediate investigation revealed that rod 30-23 cas in the fully withdrm:n position t.hile rod 26-23 was being withdraun for its friction

. test., This situation was a result of inadequate

• y ementation of administrat!ve or procedural controls and constituted a violation of Section 1. A.8 of the Technical Specifications.

Section 14.5.3.2 o,f the Vermont Yankee FSAR deals with cont rol rod withdre,eal errors when the reactor is at pover levels belou the power range. The r,ost severe case occurs when the reactor is just critical at rocu temperature and an out-of-sequence rbd is continuously withdraun.

The resultn of ther.e analyses indicate that no fuel daunge will occur ,

due to the rod withdrawal.

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The station had been in a planned shutdoun condition since September 28, 1973, in order to perform core reconstitution and interconnection of the Advanced Off-Gas System. On !!ovenber 7, 1973, work had progressed to the point where final core loading had been completed. At that point, it became desirable to perforu final core verification concurrent with control rod timing and friction tests. In order to accomodate both requirements, it was necessary to install jumpers to the refuel interlock portica of the Reactor Manual Control System in order to allow traversing of the television camera nounted on the fuel grapple chile performing control

, red friction and timing tests. Although the intent of installing the junpers was reasonabic and proper, the ensuing inplementation of this program went beyond the. scope of original intent. 'lhe reasons for this uere the inadequacy of interdepartnental contunications; in addition, certain procedures de:constrated incdequacies, specifically l'urth er, AP 504, Lifted Leads Log, OP 408, Control llod Drive System.

the control rod friction testing uns being performed in accordance uith a Startup Test Proccdure; an approved opcrating procedure did not exist.

The result of the jumper installation was a condition of interlocks thich did not prevent withdrawal of nore than ene control rod at a tire.

The operating personnel cere not adequately inferned of the jumpered interlock status; centrol rod testing was resumed concurrent with core verification. As control rod testing progressed, rod 30-23 uns inadvertantly Icft in the fully withdraun position. After core that verification cor. trol tas completed, and since the reactor operator was not cegnizant rod 30-23 was still uithd::.un, an adjacent lateral control rod 26-23 cas selected cnd its continuous withdrawal begun in preparation for the fricticr.

test. Detteen not ch posit ion O and 26, the operr. tor noticed rnpid source range conitor response. He inmediately initiatcJ control rod insertion.

At th$ t ime a full rod scram 1:ss initiated by the interr;cdiat e ran;;c conit or hi rh-high flus signals . It was later d:monstrated that contr4 rod 30-23 diy, ital position display was functioninl properly. The reactor operat or could not explain his failure to observe the indication of centrol rod 30-23 being fully withdraun.

The iruediate action of the Shif t Supervisor on duty uns to notify birher plant nanancent and to detert. ire i f pe: sonnel were on the re fueling floor during the incident and to request d:wl:4eter reading- of all perronnH at that location en the conservat ive ar.su.:pt ion that a criticality nay heve occurred. 1:ive personnel were on'lhe the refueling floor at the tiec in areas not adiecent t o t he open vennel . i;:xinne dc ;iret er reading of the personkel involved u.. 25 nr; hor.ever thir t ot al uns accunolat ed over a 11 was five hour work period nnd not at t ribut abic t o this incident alone.

also teri fied that the lec.il area nauitors, the contin w a:, ai r uc.:i t or on the refnelin:; floor, as uc11 a- the 1:eactor I:uilding Vent ilation !!.c ..ne t monit on :.hoved m incre.r.ed level of radict ion.

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,t, Page 3 Following the arrival on site of the Assistant Pla'nt Superintendent und the Reactor Engineer, further evaluation determined that the scope of installed jumpers was beyond the original intent. 'the jumpers ucre removed and it was decided to perform a subcriticality test on cach of the two involved control rods which verified their proper effectiveness.

Based upon the above evaluations, it was determined that no fuel failure had occurred and no radiation problem existed. The installed interlock jumpers were removed and a verification test conducted to determine that the rod block interlock was restored.

On Novenber 8,1973, consnitation with off-site higher management and engineering personnel resulted in the renoval of the involved fuel assemblics from the core for sipping and visual inspection. No evidence of Icakage or visual degradation was observed. The following is a listing

'of the assemblics examined and their location:

Assembly Number Core Location VT 164* 27-22

\T 171* - 29-22 VT 167 27-24 Vr 175 '29-24 Vf 049 31-32 In addition, a two rod critien1 test was conducted utilizing control rods 30-23 and 26-23. As a result of this test , it was deternined that with control rod 30-23 in the fully withdrawn position, criticality was achieved when control rod 26-23 was withdr:azn to notch 16.

The filn badr,cs assigned to personnel on the refueling floor at the tire Cf the incident ucre sent out for proc.c:.s ing. 'the recults of the j badge bearing neut ron sensing indicated a t ot aj of 50 or bet a-garena and i zero neutron exposure. This tot al badge eximure was acetoulated over a two day work period. The results of the remaining four bedres indiented that two badr.es neasured 20 nr bet a-r.aum and tuo hadges reasured 0 nr b et a- 9,amma .

Subscrpent calcul ut i ons. by Q2neral Elect ri e Co. veri fied crit ica)) t y Further c:.lculati on at not ch 16 on rod 26-23 with rod 50-23 fully uithdraun.

by Cencral 1:1cetric Co. det ermined that with red 30-23 fully withdr: n and and had rod 20-23 rod 26-23 at not ch 26, the exs cw react ivity was 0.0M t.l.,

been fu))y withdraun, the excess venet ivity would have bt en 0.!)M /J.

• These assenhlies we re visually in'.pect ed.

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. j Directorate of Licensing November 14, 1973 Page 4 General Electric personnel with recognized competency in the arca of core hinctics, and in particular control rod drop accidents, uncontrolled withdrawal incidents, etc. , did a qualitative evaluation of what transr ' red based on the above statistical information. An estimate based upon many previous calculations of a similar nature, was that the bounding results were as follows. The peak fuel center line temperature would have increased no mere than 500*F and the peak clad temperature would have-increased no more than 50 F from the starting conditions. Therefore, the fuel center line temperature was no higher than 585 F an.d the peak clad temperature was no higher than 135*F.

Plant management has discussed at length with all involved personnel

,tl)c signi icance f of this incident and stressed the areas of inadequate personnel performance. Further, a review h:is been uade of the past and present performance of the employees directly invohed in this incident.

• mis assessment has determined that these employees are capable, sincere, and conscientous and that every reasonable assurance exists that they are adequately qualified in all respects to continue in their present assigned job responsibilities.

Upon completion of an indeptl evaluation of the total' incident and the various now apparent inadequacies, it is concluded that no singular outstanding arca was predominant.

The Plant Operations Review Committec (PORC), uct to rev.iew the inci dent and made the following recommendations and/or conclusions:

I 1. The original int ent of the jumpers was reasonabic; however, the f.inal condition obtained was improper and the applied jumper::. should have been renoved imaediately following the complet i on of core veri ficat ion. ,

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2. The results obtained from the fuel asse:::blies sipped and inspected on Noven'>er 8,1973, showed no observed indicatj ons which would preclude pl mt startup.

The P1 vnt Operat i ons Revi ew Connit t ee. quest i oned wht the r adt quat e sensitivity to sipping still existed con ,j dering the elapsed shut doun tir.e and recen x aded 't al ing tua ). noun lea!,ers previon t i) reuoved during this shutdpun and sippiny, t o dat ernine if adc qunte sensitivity still existcd. On ::oveuher 11, 1973 tuo fuel I '2 l assemh)ies were sipped in an aticcpt t o prove l I3I and I

) sensitivity. Th( posj t ive J esul t: Obinined veri y the udequacy l of sipping, sensit ivit j os oh ;e rved on l orci,her C, 19 '/3 .

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2 Suberitical testing results of the two involved control rods and the management evaluation of the plant condition on November 7,1973, were deemed sufficient to permit further control rod friction testing following the incident.

4. Administrative Procedure AP 504 " Lifted Lead Log" was not adhered to. Jumper installation was not recorded in the general plant log.

5. All plant procedures relating to control rod movement shall be modified to reflect interlock requirements imposed by the reactor mode swit ch positi on.

6. Specific operating procedures addressing control rod friction and settling test s sha.11. be developed.

7. The present AP 501, Lifted Leads Log procedure, is inadequate and a PO!!C sub-committee has been appointed to review and/or revise the current procedure.

S. Until the above appointed PORC sub-committep performs its task, no installation of jumpers or Jifted Icads shall be perforced on the circuitry associated with the Reactor Protection S,vstem, the Primary Contain::.cnt Isolation Systen, any ECC System, the Reactor Manual Control System and any refuel interlock until approved by PORC.

9. No further tuo (2) rod critical testing shall be performed on side by side rods.

10. The following itens contributed to the incident:

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a. A lack of definition on the interfacing of responsjhilities on an int erdepartment al level,

b. Failure by p1m.1 supervisi on t o exercise ri norous slept icist.,

relati ve to abnornal or inadequat e plant conditions that are encountcred. ,

c. Operator error. -

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. l At the request of the I!: nager of Operations, the I?uclear Salcty Audit and Review Com..ittee ract in a special meetin;; on November 14, J973, to review the incident. The NSAR returned the following conclusiens:

1.  !!o unrevicwed safety question was involved.

2. llic health and safety of the public and plant personnel was not impaired.

3. There is no undue rish to the health and safety of the public if the plant is started up and operated in accord with the proposed schedule. .

Sincerely, VEPJ:GNT YANKEli NUCLEAR PO'.iER COPJ'ORATICS

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. Report No. 73-29 ,

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St)MEct: Violation of the Technical' Specifications, paragraph 3.8 A, l i

in that during power operation, by virtue of the fact that an  !

I inoperable snubber existed on steam Ilnes to each of the two i,

.lsolat.Lon Condensers, both conden~sers were considorod to bo

. inoperabl e. -

f This event is ccasidered to bo.an abnorcal occurren'cc as defined I in..the Technical Specifications, paragraph 1.15B and D. Notifi/

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cation of this event,'as required by the Technical Spr cifications, paragraph 6.6.2.n, viti mde to td. Region'i, Dirictorate of Regu-latory Operations,-by telephone on $sturdsy, Nover.her 3,1973, at

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,, l 0850, and by telecopior on Honday, Ndvether $,1973, ht 1315. 1 g.

~.l SITUATION: h* nile conducting in 'inipection of the hydraulic shoch'and sway l

!streM. ors (snuhbers) located on varl us systens-in the' Reactor, Building, but. outside of the Drheell, the accteulators on one.  ;

i unit'on the steam line to the A IsN otion Condenser bnd one unit ,

on the staan line to the B Isolation Condenser were found to be l

t devoid of fluid, soth units we2+ considered to be inoperabic. l l

To bo detemined upon inspection. .

CAtlSE:

l ItEHRDIAL Af7 TON: -

l As per the requfrements of the Technical Specifications, para-graph 3 d' ,0, an orderly plnnt shutdown was comenced upon noti-4 THIS DOCUMENT CONTAINS  !

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e fication of the situation at ' 1850. 14canwhile, imedinte. efforts

- were cade. to refill' t.h'c sntthbor accumulator on the A Isolation Condenser atsam line, 'Ihis action was completed by 1845, re- -

turning the sm6ber"to ' service, The load drop which had been started was halted snu autput was.again incrensed to the initial-

' level, Follow-up action ' included replacement of.the acntmaintor on the snubber installed on the Ti Isolation Condensor stecn line, then replacement Of the entire.spubber unit' on the A Isolation condenser steam line. Tnis acti'on was corp 1sted by 1910 Friday cycning. A follow-up check was then made on Saturday evening to insure that'no fluid Io.fs proh' i?. existed, SAFHn' SHNIFICANCE:

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Atendrent 67 t'o 'the EDSAR detailr. the,requirepnts .for at losst i, -

one Isolation Condenser to be available as a .hect sink in the i

cycnt of a li)s.iof Coolt.nti-Accident. . In this.sittution, it can he postulated,that this requirement r.ight not have'been r.et, had an earthqtiske . occurred which would require .the snubber to be fully 'openthie, E

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