ML20148U134

AO 73-31:on 731107,reactor Scram Occurred During Control Rod 26-23 Friction Test.Caused by Inadequate Implementation of Administrative or Procedural Controls.Control Rod 30-23 Was Fully Withdrawn,While 26-23 Was Being Withdrawn

ML20148U134
Person / Time
Site:	Vermont Yankee
Issue date:	11/14/1973
From:	Riley B VERMONT YANKEE NUCLEAR POWER CORP.
To:	US ATOMIC ENERGY COMMISSION (AEC)
References
AO-73-31, VYV-3071, NUDOCS 8103020770
Download: ML20148U134 (6)

Text

.A. . , _ , , , , _ , _ _ ._ . _ , _

m-e-- =s < d. ym ,,. .. ,_ . .

a

g. ,'-

l

%. b g

SCVcNTY SCVCN GHOVC STRCCT RurLAxo, Vt:im'ONT 0 5'/01 HCPLY TO VYV-3071 e. o. uox 157 4ERNON. VERMONT o'M .54 November 14, 1973 ,

i Director Directorate of Licensing United States Atomi c Energy Commission h'ashington, D.C. 20545

REFERENCE:

Operating License DPR-28

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

As defined in Section 6.7.B.1 of the Technical Sjiccifications for the Vermont Yankee Nuclear Power Station, we are reporting the following Abnormal Occurrence as A0-73-31. -

On November 7, I973, at 2101, while the plant was in a shutdown condition and while the reo,uired Control Rod Friction testing was being perforned on control rod 26-23, a reactor scram otenrred initiated by a high-high flux signal from the Intermediate Range Neutron Monitoring System.

An immediate investigation revealed that rod 30-23 was in the fully withdraun position while rod 26-23 was being withdrmen for its friction

. test., This situation was a result of inadequate implementation of administrative or procedural controls and constitated a violation of Section 1. A.8 of the Technical Specifications. l l

I Section 14.5.3.2 o.f the Vermont Yankee FSAR deals with control rod withdrawal errors when the reactor is at power IcVels below the power range. The most severo case occurs when the reactor is just c2 i ti ca l at room temperature and an out-of-sequenc:e rod is continuously wi* hdrawn.

The results of these analyses indicate that no fuel damage will o eur due to the rod withdrawal, ,

.THIS DOCUMENT CONTAINS t P00R QUAUTY PAGES , ,

8/0303f 0 '

-_ y_ _ _ - _ _ ____

m, ,

_ _f _ _ _ _ _ _ _ _ _o...

, ,J ~ 4.,  ;

Y a... _ . {dL ' , .x o .

9. d w x w x .

x I

r; '! .

k 1 "n ,

VcnMONT YANKCC NUCL.CAfi r*OWER CORPORA *f f 0

,n s, 3

.\

_t.

ll .

iDirectorate'of Licensing ~j

' .s

' M November ' 14, 1973 '

.Pago 2.'

i j . 7the' station. had been in n' planned shutdown condition s nce September -28, .1973, in order to perform core reconstitution and

is bonnection of the Advanced Off-Gas. System. On November 7, -

U 1973, wolk had progressed to the point where final core loading

' had been : coupleted. At that point ,. it' became desirabic to, perform .1 final core verification concurrent with. control rod timing and friction' tests. In order to accomodate both reqairements, it was E inecessary. to-install. jumpers to the refuci interlock portion of the Reactor' Manual Control System in order .to allow traversing of the. 1 television camera mounted on the fuel grappic while' performing control 1, rod friction and timing tests. . Although the intent of installing

.the jumpers was reasonabic and proper, the ensuing impicmentation of

- this pronram went beyond the scope of original ~ intent. 'lhe reasons i

' for this were the inadequacy of interdepartmental communications; in addition, certain procedures demonstrated inadequacies, specifically

= AP 504,1,if ted Leads. Log, OP 408, Control llod Drive System. . Further,

'the. control rod friction testing was being performed in accordance with p' a Startup. Test'Procedurc; an approved operating procedure did not exist.

' Ihe result of the jurper installation was a condition of interlocks.

((

E j which:did not prevent withdraual of nore than one control rod at a time.

Th'c operating personnel were not adequately inforned of the jumpered .]

' interlock status; control. rod testing was resumed concurrent with core {

i

. veri fi cati on . As control rod testing progressed, rod 30-23 was After core verification inadvertantly Icft in the fully withdrawn position.

was completed, and since the reactor operator was not cognizant that control j rod 30-23 was still withdraun, an adjacent lateral contro! rod 26-23 was j selected and its continuous withdrawal begun in preparation for the friction test. Detween notch position 20 and 26, the operator noticed rapid source range monitor response, lie immediately initiated control rod insertion.

At thjt tinc a full rod scram was initiated by the . ntermediate

~

i runhe i

)

monitor hi gh-high. flux. si gnals . It was later de aonstrated that control rod 30-23 digital position display was functioning properly. The reactor operator could not explain his failure to observe the indication of control rod" 30-23 being fu'))y withdrawn.

The immediate action of the Shift Supervisor on duty was to notify higher' plant canagement and, to det ermine if personnel were on the refueling l

. floor- during the incident and to rgquest desincter readings of all perronne k, at that lo':ation on the conservative ansubplan that a criticality may heve L

. occurreC Five personnel were on the refueling floor at the time in areas not adjacent t o t he open vessel . The maxinum dosineter reading of the h' pers6nnel-involyc0 was 25 nr; however, thir total was accouplated over a 11 W3S

. five hour work. period and not 311 T.il'ut able 1 o t hi S i n e.i de nt alone.

air uenitor on alho; veri [Ied 'that the lecal area MOHitDTs , the cont inhouS the refuelinh floor, as well an the Henetor I:uilding i'ent ilation fixhaust - '

monit ors showed no increar.ed le m) of radinfion.

!L

( ,

L_ _; e l i ,+ . ; 3. d. .

, J.^ : -- . .

(

t i 4 1 ,

• VERMONT YANKCC NUCLCAR POWCR CORR'OMAT:C y;

.t-

\~ *

'Directoratciof Licensing 3' '

? November 14,;1973 '

y Page.3 t

d

Fo11' o wing the arrival on site of the .Ansistant' Plant Superinten ent and the Reactor Engineer, furthet cvaluation determined that the' scope of' installed jumpers was beyond;the origina1Lintent. 1hc 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 determin,d that no fuel failure

-had ' occurred 'and no radiation problem existed. The installed interlock. -'

. jumpers were removed and u. verification test conducted to determine that the rod block interlock was restored.

On Novenber 8,1973, consultation with off-site higher. management and engineering personnel resulted in .the removal of the . involved fucI

• 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 i

VT 164* 27-22 ,

- Y1' 171 * - 29-22~

.VT 167 27-24 , '

.VT 175 29-24 Vf 049 31-32 In addition, a two rod critical 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 withdrain to notch 16.

The film badges ansigned to personnel on the refueling floor at the time M the incident ucre sent out for proecssing. 1hc results of the ,

badge bearing neutron senning indicated a t ot al of 50 mr bet a-garna and zero neutron exposure. This tot al badge exposure wa's accomlated over a two day wor 1. period. The results of the remain. inn, four badges indicated that two badr.cs measured 20 mr beta-gamma and tuo badges neasured 0 mr .; .

bet a- p.amda .

Subsequent ; calculat ions by General E1cet rie Co. ver.i fled crit i cali t y at not ch 10 on : rod 26-23 Wi1h rod 50-23 fully ei1hdraun. Further ca) cult. tion

-by General Elect ric Co. dct ermined t hat with red 30-23 fully withdre.cn and rod 26-23 at. notch 26. the excess reactivity was 0.07% /1, and had rod 20-23 l'

~ been' fully T:i thdrawn, t he cxcess react ivity would have been 0.970 4K. >

l.

f -

l t .

'

• These. assenblies were visually i ntpect ed. ,

3 J

. l

,y: w A y- - . . .

---

.--a..~

- _.. i .: ..

.l t.' ,.

c. . ,

l' . .s - 4 .

VERMONT YANKEC NUCLCAR POWCN CORI'OR AT: .*. 1 a .  ;

O Directorate of Licensing -

November.14, 1973.

- Page 4 .

1

• 1 l

General Electric personnel with recognized competency in the area 'of core kinetics, and. inL particular control rod drop. accidents , uncontrolled withdrawal incidents,. ctc. , did a qualitative. evaluation of what transpired based .on the aboycL statistical information. An estimate based upon many

- previous calculations. of. a- similar nature, was that the bounding results

- werelas follows. , The peak fuel- conter.line temperature would have  :

increased ^no more than 500*F and the: peak clad temperature would have  :

increased no more than 50 F-from the starting conditions. Therefore, the fuel centerc line temperature l was no higher than 585*F and the peak clad i temperature was no higher than 135*F.

~ Plant management.has discussed at length.with all involved personnel the significance of this incident.and stressed the arcas of inadequate personnel performance. Further, a review has been nade of the past and-present performance of the employces directly' involved in this incident.

Thisssessment a

~

htis determined' that tijese employees are capabic, sincere, and conscientous and that crory reasonable assurance exists-that they are

~

adequately qualified in al1~ respects to continue in their present assigned

- job responsibilities.

Upon completion of an indeptl$ evaluation of the total' inci, dent and

, tho various now apparent inadequacies, it is concluded that no singular outstanding arca was predominant.

The Plant Operations lieview Committec (PORC), uct to review the incident and made the following recommendations and/or conclusions:

1. The original int ent of the jumpern was reasonnble; however,

-the final condition obtained was improper and the applied jumpers should have been removed immediately following the comnletion 'of core verificalion. ,

j

-r

2. The results obtained from the fuel assemblies sipped and inspected on Novei.her 8,1973, showed no observed indications j

-which would preclude plant startup.

The Plant Operati ons McVi ew C(n.'.mit t ee quest ioned t;hether adequat e ,

sensitivity to sipping still c.sinted considering the elapsed  !

shutdoun tira and recoauended t aking tuo knoien leal,ers previum.ly reuoved during this shutdoun and sippiny. t o dat eruine if adequnt e '

sen5.Itivily Still e.\ibled. On IovemhcT 1*l, 1973. tu0 fuel '

assemblics were si; sped in an at tempt to prove 1 131 and 1 132 4 ' sensitivity. The positive jesu11a obtainql verify the adequacy of. s' pping sennit ivit ics ohne rved on !!oveLber 8, 1973.

. I s' [

1 s -4 . w . . L .:, , - , .

_- / / A .- ~ . - , .

l, ,, VCRMONT YANKCE NUCLt2\n POWER CORl'OR.V;'.c.

3 3

Directorate of Licensing i November 14, 1973 Page 5 4

{

3. Suberitical testing results of the two involved control rods and the management evaluction of the plant condition on November 7, l'J73, were deemed suff.icient to permit fu rt h er 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 reficct interlock requirements imposed by the reactor mode switch position.

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

7. The present AP 501, Lifted Leads Log procedure, is inadequate and a PO!1C sub-comittee has been appointed to rev.iew and/or revise the current procedure.

8. Until the above appointed PORC sub-comm.itt es performs its task, no innt.n)]ation of jua.pers or lifted 1 cads shall be perforued on thi: circuitry associated with the Reactor Protection System, the Primary Contair.::.cnt Isolation System, any ECC System, the Reactor t!anual Control System and any refuel interlock until approved by PORC.

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

10. The follerine,- ite: contributed to the incident ;

co

a. A lack of definition on the interfacing of responsjhilities on an interdcparunent al level.

$ b. Failtire by plant supervision t o exercine rigorotu slept i ci r:

relat ive to abnorm 1 or inadequate plant conditions that are encount eveJ.

c. Operator error -

l

1. . .= , , , r . - . . . - . . . .

LL . , .. ..s . + '

. }

• i

, /.

's VCRMONT YANKCC NUCLEAR POWER CORPOR,VT :c '

Directorate of Licensing November 14, 1973 ,

Page 6 At the request of the Manager of Operations, the Nuclear Safety Audit and Review Committee met in a special meeting on November 14, 3973, to review the incident. The NSAR returned the following conclusions:

1. No unreviewed safety question was involved.

2. The 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, YERMOST YANKEE NUCLEAR POWER CORPORATICS

, 'O s bU c=. -1 O B . l'! . Ri l ey Plant Superintendent P,WR/llFC/Lbd .

(p i

s l

.