ML20148F751

Forwards Bulletin 73-06, Inadvertent Criticality in Bwr. Action Required

ML20148F751
Person / Time
Site:	Yankee Rowe
Issue date:	11/27/1973
From:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Minnick L YANKEE ATOMIC ELECTRIC CO.
References
NUDOCS 8011060720
Download: ML20148F751 (1)

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j Docket No. 50-29 l Yankee Atomic Electric Company-  ;

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• The enclosed Directorate of Regulatory Operations Bulletin No. 73-6, l d " Inadvertent Criticality in a Boiling Water Reactor" has been sent for completion of the requested action to utilities that are presently licensed

(or that will be licensed in the near future) to operate boiling water reactors.

This bulletin is being sent to you to provide you with information f j that was reported

• to the AEC by the Vermont Yankee Nuclear Power Corporation, concerning an inadvertent criticality incident that was experienced at their Vermont Yankee facility. This bulletin is provided only for your h@ ,

general information. ,

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RO Bulletin No. 73-6 3 November 26, 1973 2

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INADVERTENT CRITICALITY IN A BOILING WATER REACTOR

$ We recently received an abnormal occurrence report from the Vermont

! Yankee Nuclear Power Corporation relating to an iriadvertent criticality

'j incident that was experienced at their Vermont Yankee facility. A copy e of the abnormal occurrence report is attached to this Bulletin to provide j , you with pertinent details of this event.

a i At the time of the inadvertent criticality incident, the reactor vessel i and primary containment heads were removed, the refueling cavity above 4 the reactor vessel was flooded, control rod friction tests were in 3 progress, the rod worth minimizer was bypassed, and core verification j- had been in progress. As a result of the incident, no measurable '

radioactivity was released, no fuel damage resulted and no personnel

. , exposures were experienced. The incident is currently under review and evaluation by the Regulatory Staff.

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- Action requested by this bulletin is contained in Section A.

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3 A. Action Requested by Licensees In light of this occurrence.you are requested to take the following actions. Upon completion of these actions you are requested to inform this office in writing, within 45 days of the date of this letter, of the status of each item identified in each paragraph and subparagraph listed below: ,

1. Procedural Review

a. Control Rod Drive Operating and Testing Procedures (1) Conduct a review of your control rod drive t.cerating '

and testing procedures to determine that approved procedures

. exist for all operations and tests.

t (2) Verify that appropriate prerequisites are included in the procedures to require testing of associated interlock and protective features before control rod testing is ,

.f permitted.  :

(3) Assure that prerequisites and detailed instructions are  !

provided that demonstrate compliance with technical spe-  !

cification requirements and design bases.  ;

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b. Bypass Installation Procedures (Jumpers or Lifting of Leads)

Assure that existing bypass installation procedures have been ,

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conservatively reviewed for technical adequacy and for adminis-strative controls.

c. Radiation Protection Procedures Assure that procedures for access control and personnel accountability in areas subject to accidents are current.

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d. Shif t Transition Procedure (Turnover) l l

Assure that complete and detailed procedures are in effect l

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• that provide instructions for a proper and conservative turnover of shift responsibilities. Such procedures must include requirements for commun,1cating the status of all safety related equipment and conditions.

p l 2. Management Controls Assure that your management controls that are in effect provide E

for qualified technical and administrative reviews and approvals of temporary circuitry changes and temporary off-normal plant conditions. This review should assure that the responsibilities and requirements associated with the review and' approval, installation, verification, removal, and subsequent testing.of temporary circuitry changes and temporary off-normal plant conditions are clearly delineated in station procedures, are understood by the station staff, and are being properly implemented.

3. Licensed Operator Performance Assure that management provides the necessary opporturities and time so that operators are adequately trained to carry out

- their assigned responsibilities. In particular, confirm that shift crew me=hers are provided special training for safety related activities that are infrequent, complex, or have unusual safety significance.

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If you have any questions concerning this request, please contact this oL'fice. ,

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Attachment:

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Vermont Yankee A0 No. 73 Letter dated November 14, 1973'to the Directorate of Licensing, USAEC, Washington, D.C.

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• i RUTLAND, URMUNT 03701 REPLY TOs VYV-3071 p, o, oox is7 VERNON. VERMONT OS354 November 14, 1973 f t

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Directorate of Licensing a.__..

United States Atomic Energy Commission .

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REFERENCE:

Operating License DPR-28 Docket No. 50-1:1 Abnormal Occurrence No. A0-73-31 Gentlemen:

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

On November 7, 1973, 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 occurred initiated by a high-high flux signal from the Intermediate Range Neutron Monitoring j System.

An immediate investigation revealed that rod 30-23 was in the fully withdrawn position while rod 26-23 was being withdrawn' for its friction test., This situation was a result of inadequate impicmentation of

.,dministrative or procedural controls and constituted a violation of Cection 1. A.S of the Technical Specifications.

l Section 14.5.3.2 of the Ver.aont Yankee FSAR deals with control rod withdrawal crrors when the reactor is at poh'cr levels below the pcwcr L

range. The most severo case occurs when the reactor is just critical ~

at room temperature and. an out-of-sequence rod is continuously withdrawn.

The resultt of these analyses indicate that no fuel damage will occur due to the .od withdrawal. ,

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'Ihc station h'ad 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, work had progressed to the point where final core loading ,

had been completed.- At that point, it became desirable to perform  ;

• final' core verification concurrent with control rod timing and. '

{. friction tests. In order to accomodato both requirements, it was .

p necessary to install jumpers to the refuel interlock portion of the p Reactor Manual Contro1' System in. order to allow traversing of the

[ television camera mounted on the fuel grapple while performing control rod friction and timing tests. Although the intent of installing 5 .

the jumpers was reasonabic and proper, the ensuing impicmontation of this program went beyond the scope of original intent, 'Ihe reasons for this were the, inadequacy of interdepartmen,tal communications; in <

addition, certain procedures demonstrated inadequacies, specifically AP 504, Lifted Leads Log, OP 408, Control Itod Drive' System, Further, .

the control rod friction testing was being performed in accordance with ,

a Startup Test Procedure; an approved operating procedure did not exist.

The result of the jumper installation _.was a condition of interlocks which did not prevent withdrawal of nore than one contrp1 rod at a time. ,

, The operating personnel were not adequately informed of the jumpored interlock status; control rod testing was resumed concurrent with core verification. As control rod testing progressed, rod 30-23 was inadvertantly Icft in the fully withdrawn position. After core verification  ;

was c,onipleted, and since the reactor operator was not cognizant that control ~

rod 30-23 was still withdrawn, an adjacent lateral control rod 26-23 was solccted 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, lic immediately initiated control rod insertion. ,

At thp timo n. full rod scrau was initiated by the intermediate range '

monitor high-high flux signals. It was later demonstrated 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 fully withdrawn. -

The immediate action of the. Shift Supervisor on duty was to notify higher' plant management and to determine if personnel were on the refuchng floor during the incident and to request dosincter readings of all personnel at that location on the conservativo assumption that a criticality may have i occurred. Five personnel'wcre on the refueling floor at the tiac in areas not adjacent to.the open vessel. The maximum dosimeter rending of the  !

' personnel involved was 25 mr; however, this total was accumplated over a '

five hour work period and not attributabic to this -incident alone. It was ,

also verified that the local area nonitors, the continuous air monitor on  ;

the refueling floor, as well a . thc. I:cactor Cuilding Ventilation lixhaust  ;

monitors: showed no inercased level of radiation. -

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. Directorate of Licensing 3y November 14, 1973 -

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Following the arrival on site of the . Assistant Pla'nt Superintendent and the Reactor Engineer, further evaluation determined that the scope

of installed jumpers was beyond'the original intent. 'lhe jumpers werc l

removed and it was decided to perform a subcriticality test on each of the two involved control rods which verified their proper effectiveness.

Based upon the above evaluations, it was determined that no fuel failurc had occurred and no radiation prob 1cm existed. 'lho installed interlock l -

jumpers were removed and a verification test conducted to determine that

) the rod block interlock was restored.

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

assemblics .from the core for sipping and visual inspection. No evidence b of Icakage or visual degradation was observed. The following is a listing

(. 'of. the assemblics examined and their location:

Assembly Number Core Location -

\T 164* 27-22

\T 171* 29-22

\T 167 27-24 Vr 175 ~

29-24 17 049 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 detornined that with control rod 30-23 in the fully withdrawn position, criticality was achieved when control rod 26-23 was withdrawn to notch 16.

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The filn had;;cs assigned to parsonnel on the refueling floor at the timo M the incident ucre sent out for processing. The results of the-badge bearing neutron sensing indicated a total of 50 mr beta-gamma and zero neutron exposure. This total badge exposure was accuwilated over a p

j two day work period. The results of the remaining four badges indicated L that two badges measured 20 nr beta-gamma and two badges measured 0 mr beta-gamma. ,

Subsequent calculations by General Electric Co. verified criticality n

at not ch 16 on rod 26-23 with rod 30-23 fully withdrawn. l'orther calculation

' by Cencral F.lectric Co., determined that with rod 30-23 fully withdrawn and rod 26-23 at notch 26, the excess reactivity was 0.67% AK, and had rod 26-23 been fully withdraun, the exccr.s reactivity would have been 0.970 AK.

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• These assemblics were visually inspected. .

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General Electric personnel with recognized competency in the area'of core hinctics, 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 more than 500*F and the peak clad tenperature 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 the significance of this incident and stressed the areas of inadequate personnel performance. Further, a review has been ma'de of the past and present performance of the employees directly involved in this incident.

,This assessment has determined that these employees are capabic, sincere, and conscientous and that every reasonabic assurance exists that they are adequately qualificd in all respects to continue in their present assigned job responsibilities.

Upon completion of an indepth evaluation of the total inci. dent and the varicas now apparent inadequacies, it is concludcnl that no singular outstanding area was predominant.-

The Plant Operations Review Committee (PORC), uct to review the

, incident and made the following recommendations and/or conclusions:

1. The original intent of the jumpers was reasonabic; however, the final condition obtained was improper and the applied jumpers should have been removed immediately following the

, completion of core verification. ,

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2. The results obtained from the fuel assemblics sipped and inspected on November 8,1973, showed no observed indications which would preclude plant startup.

The Plant Operations Review Committee. questioned whether adequate sensitivity to sipping still existed considering the clapsed shutdown time and recomucnded taking two known Icahers previous]y removed during this shutdown and sipping to deteruine if adequate sensitivi ty still existed. On November 11, 1973, two fue) assemblics were sipped in-an attempt to prove 1 131 and 1 132 sensitivity. The positive results obtained verify the adequacy of sipping sensitivitics observed on November 8,1973.

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L 3. Suberitical testing results of the two involved control rods g: and the management evaluation of the plant condition on L

" Noveber 7, '1973, were deemed sufficient to permit further gn control rod friction testing following the incident.

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4. Administrative Procedure AP 504 " Lifted Lead Log"' was not g4l adhered to. Jumpor installation was not recorded in the di

• general plant log.

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g All plant procedures re'lating to control rod movement shall be fp modified to reficct interlock requirements imposed by the reactor modo switch position. .

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+ ' settling test s shall. be developed.

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The present AP 504, Lifted Leads Log procedure, is inadequate NN and a PORC sub-committee has been appointed to review and/or revisc the current procedure.

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e Until the above appointed P. ORC sub-committe.e performs its . task, W no installation of jumpers or lifted Icads shall be performed on the circuitry associated with the Reactor Protection System, the Primary Containment Isolation System, any ECC System, the Reacter Manual Control System and any refuci ' interlock until approved by PORC. .

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

10. The following ' items contributed to the incident; er

a. A lack of definition on the interfacing of responsibilitics on an interdepartmental leycl.

b. Failure by plant supervision to exercise rigorous skepticism l

relative to abnormal or inadequate plant conditions that are' encoMntered.

c. Operator crror.

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NUCLEAR POWER CORPORATI Directorate of Licensing- ,.

November'14, 1973 * '

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At the request of the Manager of Operations, the nuclear Safety 7 ,

.Aud t and Review Conunittec met in. a special meeting on November 14, i

1973, to review the -incident. The NSAR returned the following

- conclusions: -

y. 1. No unreviewed safety question was involved. .

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2. The health and safety of the public and plant personnel '

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.t not impaired. -

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[ 3. Thero' is no -undue risk to tho' health and safety of the public y' if the plant'is started up and operated in accord with the 1 proposed schedule.

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p YERMONT YANKEE NUCLEAR P0h'ER CORPORATION ,

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. B.tl. Riley Plant Superintendent Bh'R/h'FC/kbd .

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