ML20086D739

Updated AO 73-31:on 731107,reactor Scram Occurred,Initiated by high-high Flux Signal from Intermediate Range Neutron Monitoring Sys.Rod 30-23 Fully Withdrawn & Rod 26-23 Being Withdrawn for Friction Test

ML20086D739
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/14/1973
From:	Riley B VERMONT YANKEE NUCLEAR POWER CORP.
To:	US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20086D716	List: ML20086D718 ML20086D739 ML20086D899
References
A-731114, AO-73-31, VYV-3071, NUDOCS 8312020480
Download: ML20086D739 (6)

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REFlilll!NCE: Operating License DPR-28 #3"far Docket No. 50-271 # -;\

Abnormal Occurrence No. A0-73-31 0 Gentlemen:

As defined in Section 6.7.11.1 of the Technical Specifications for the Vermont Yankee Nuclear Power Stat ion, we are reporting the following Abnormal Occurrence as A0-73-31. .

On Nover.ber 7, 1973, at 2101, while the plant was in a shutdown condit ion and while the requi red Control Rod Friction testing was being perforned on control rod 20-13, a reactor scrau occurred initiated by a hip,h-high flux si ;nal 3 fron the Intermediate Range Neutron I!onitoring System.

An inmediate investigation revealed that rod 30-23 was in the fully wi t hdraun pos it i on while rod 26-23 was being withdraunifor its friction test., This sit uation was a result of inadequat e juplementation of administrative or procedural controls and constituted a violation of-Sect ion 1. A.8 of the Technical Speci ficat.lons.

Section 14.5.3.2 of the Vermont Yankee FSAR daals with control rod withdrawal errors when the reactor is at poNer levels below the power range. The most severe case occurr, when the renetor is just critical at room temperature and an out-of-sequence rod is continuously withdrawn.

The results of these analyr.es indicate that no fuel damage will occur' due to the rod withdrawal, ,

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November 14, 1973 s Page R l .

The station had been in a planned shutdown condition since September 28, 1973, in order to perform core reconstitution and interconnection of the Advanced Off-Gas System. On November 7, 1973, ucrk had progressed to the point where final core loading had been completed. At that point, it became desi rable to perform _

final core verification concurrent with control rod timing and friction tests. In order to accomodate both requi rements , it was necessary to install jumpers to the refuel interlock portion of the Itcactor Manual Control System in order to allow t raversing of the television camera mounted on the fuel grapple while performing control -'

rod friction and timing tests. Although the intent of installing the jumpers was reasonable and proper, the ensuing implementation of this program went beyond the scope of original intent. 1hc reasons '

for this ucre the inadequacy of interdepartmental con.munications; in

! addition, cert ain procedures dea.onst rated inadequacies , specifically 4 AP 504,1,i f ted Leads 1.og, OP 403, Cont rol 1:od Drive System, ru rth er, the control rod friction testing was bring performed in necordance with-a St artup Test Proccdure; an approved operating procedure did not exist.

i The result of the jumper installation was a condit ion of interlochs which did not prevent withdraual of nore t han one-control rod at a time.

The operating pe rsonnel were not adequatejy informed of the jumpered interlock stat u: ; control rod testing was resumed concurrent with core ve ri fi ca t i on . As control rod testing progressed, rod 30-23 was inadvertantly left in the fully withdraun position. After core verification was con.pleted, and since the reactor operator was not cognizant that control rod 30-23 was sti11 withdrawn, .m adjacent lateral control rod 26-23 was selected and its continuous withdrawal hegun in preparation for the friction test. Uetween notch position 20 and 26, the operator noticed rapid source range monitor response. Ile innediat ely initiated cont rol rod insertion.

At thjf. t ima a full rod scram was init inted by the. intermediat e range monitor high-hinh flux signals. It was later denonstrated that control rod 30-23 digital positien display was funct ion ing Ji mperly. The reactor operator could not explain his failure t o observe the indication of control rod 30-23 heing fully withdr:nen.

The immediate action of the Shi ft Supervisor on duty was to notify I higher plant manancunt and to deternine i f personnel were' on the refueling floor during the incident and t o request dosimeter readings of all personnel at that location en the conservative awumption that a crit icality may have occurred. Five p"rsonnel were on the refueling floor at the time in areas not adjacent t o the open ve: sel. 1ho h :inum dosincter reading of the

[ personnel involved uas 25 nr; however, this total was acenuplated over a i five hour work period and not at t ribut able to this incident alone. It was also vc ri fied that the leest area noni tors , t he cent inuous ai r menitor on the refueling floor, as cell ". the I:cact or liuilding- Ventilat ion Exhaust mon i t o r showed no increawd level of radiat ion. -

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. VCRMONT YAjEC NUCLEAR POWCR CC4RT' ORATION 1

Directorate of Licensing

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November 14, 1973 Page 3 e, Following the arrival on site of the Assistant Pla'nt Superintendent and the Iteactor Engineer, further evaluation determined that the scope of installed jumpers was beyond the original intent. The jumpers were removed and it was decided to perform a suberiticality test on each of the two involved control rods which verified their proper effectiveness.

l'ased upon the above evaluations , it was determined that no fuel failure had occurred and no radiation problem existed. The installed liiterlock jumpers were removed an 1 a verification test conducted t o determine that the rod block interlock was restored.

On November 8,1973, consultation uith off-site higher management and engineering personnel resulted in the removal of the involved fuel assemblies from the core for sipping and visual inspection. No evidence of lenhage or visual degradation was observed. The following is a listing of the assemblies examined and their location:

g embly Number Core 1.ocation VT 164* 27-22

\T 171* 29-22 VT 167 27-24 Vf 175 29-24 VT 049 31-32 In addit ion, a two rod critical test was conducted utilizing control rods 30- 23 and G- 23. As a result of this test, it was deternined that with cont rol rod 50-23 in the fully tithdrawn position, criticality was achieved when cont rol rod 26-23 was withdrawn to notch 16.

The fiin badges assigned to persennel on the refueling iloor at the thce of the incident were sent out for processing. 'lhe results of the badge bearing neut ron sensing indicated a t otal of 50 nr beta-c,mma and zero neut ron exposure. This t v al badge c:.isosure was necunulat ed over a two day worl; period. The results of the remaining four badges indicated that two badges neasured 20 nr beta-gamn. and two bak,es measured 0 mr he t a- v.au:n .

Subsequent calculat ions by General Elect ric Co. verified criticality at not ch 16 on rod 26-23 uith rod 30-23 ful!y wi thdiaen. Further calculation by General Electric Co. determined that with rod 30-23 fully withd emsn and rod .4 23 at notch 26, 'ho es.cew reactivi t y was 0.07, AK , :md had rod 26-23 been fully withdraien, t he excess react ivity would have been 0.97', AK.

• These ' scrahlics were visually inspected. .

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VERMONT Y EC NUCLEAR POWCM CORPORATION Directorate of Licensing November 14, 1973 Page 4 General !!1ectric personnel with recognized competency in the area of cere kinetics, and in particular control rod drop accidents, uncontrolled withdrawal incidents, etc. , did a qualitative evaluation of what transpired 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 nore than 500*P and the peak clad temperature would have j increased no more than LO*F from the starting conditions. Therefore, the fuel center line temperature was no higher than 585"F and the peak clad temperature was no higher than 135*F. ,

I Plant nanagement has discussed at length with all involved personnel the significance of this incident and stressed the arcas or inadequate personnel performance. Further, a review has been nade of the past and present performance of the enployees directly involved in this incident.

'this assessnent bas deternined that these employees are capable, sincere, i and conscientous and that every reasonable assurance exists that they are i adequately qualified in all respects to continue in their present assigned '

job responsibilitics, t

llpon con.pletion of an indepth evaluation of the total incident and the various now apparent inadequacies, it is concluded that no singular outstanding area was predoainant. g The Plant Operations heriew Committee (PORC), uct to review the I incident and nade the following recomendations and/or conclusions: f I

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1. The original intent of the jumpers was reasonnble; however,  !

the final condition obtained was inproper and the applied -

j umpe r: should have been rcroved immediately following the

, complet ion of core verification. ,,

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2. The results obtained frou the fuel assemblies sipped and i inspected on Neven5cr 8, 1975, showed no observed indications l uhich would prceluJe plant startup. -

The Plant Operations Revicu Coumittec questioned whether adequate f' sensitivity to sipping still existed considerino, the elapsed shutdo':n time and reconnded taking two known lenkers previously I reuoved during this shutdoen and sipping to deteruine if adequate l sensitivity ati11 existed. Cn Novenhe r 14, 19'/3 two fuel  !

assenhlics were sipped in an attempt to prove l l31 and I IN [

sensitivity. The positive 1esults obtained verify the adequacy ,

of sipping sensit ivitii.s observed on November 3, 1973.

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Directorate of Licensing i November 14, 1973 Page 5 }

3. Subcritical testing results of the two involved control rods and the management evaluat ion 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. Ju' aper installation was not recorded in the general plant log.

5. All plant procedures relating to control rod movement shall be modified to reflect interloch requirenents imposed by the reactor node switch position.

6. Specific operat inn procedures addressing control rod friction and set tling test s shall. be developed.

7. The present AP 504, Lifted Leads Lor; procedure, is inadequate and .i PORC sub-co=nittee has been appointed to review and/or revise the current procedure.

8. Until the above appointed PORC sub-co=nittep perforns its task, no installatien of jumpers or lifted leads shall be perforued on the circuitry associated with the Reactor Protection System, the Primary Contalment Isolation Systru, any ECC System, the Reactor !!anual cont rol Systen and any refuel interlock until app roved by l'URC.

9. No further two (3 rod critical testing shall be performed on side by siA rods.

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'lhe follocine items cont ributed to the incident ;

a. A lack of definition on the interfacing of responsibilities on an interdcpartment al level.

b. 1:ailure by plant supervision to exercise rigorous skepticisu relative to abnormal o: inadequate plant conditions that are encountered.

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c. Ope rator e rror. . l l

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Directorate of 1.icensing .l Noven:ber 14, 1973 i l Page 6 j

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At the request of the Manager of Operations, the Nuclear Safety Audit and lievieu Coramittee net in a special necting on November 14, 1973, to review the incident. The NSAR returned the following  ;

conc lin. ions :

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1. 1:o unreviewed safety question was involycd.

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

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3. There is no undue rini, to the health and safety of the public j if the plant is stnrted up and operated in accord with the [

proposed schedule.  ;

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Sincerely, l 1 -

VliPJ:051' YANKlili i llCl. liar PON!!P, CORPORATION '

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11. ll . Riley '

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