ML20024B209

Reviews Current Events - Power Reactors, Mar-Apr 1976.No Action Required

ML20024B209
Person / Time
Site:	Crane
Issue date:	08/20/1976
From:	J. J. Barton, Toole R GENERAL PUBLIC UTILITIES CORP.
To:	GENERAL PUBLIC UTILITIES CORP.
References
TASK-*, TASK-GB GPU-2484, NUDOCS 8307070334
Download: ML20024B209 (16)

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AEC DOCUMENT REVIEW

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Plant / Unit _J!L-

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The attached AEC document has been reviewed for test program and design modification requirements for the above Plant / Unit.

DOCUMENT:

Operating Experience, dated:

@ Current Events - Power Reactors, dated:$46?V-86/_ /976f,j/gf,,gsg Other dated:

Review of the attached document has concluded that no action is' required.

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artup & fest Manager Date V /d R 7to 4

Test' gerintencent Date Review of the attached document has concluded that action is required by:

Problem Report (s) j has/have been issued.

Startup & Test Manager Date Test Super 1ntencent Date W

06995 DISTRIBUTI0ti:

R.W. Heward, Jr.

W.T. Gunn E.D. McDevitt' J.E. Xunkel M.A. Nelson R.J. Toole J.T. Faulkner File 416 8307070334 760820 PDR ADOCK 05000289 S

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!!E UNITED STATES CURRENT EVENTS NUCLEAR i

REGULATORY a

POURR REACTORS commission i

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EVENTS SELECTED FROM REFORTS SU5MITTED TO THE UNITED STATES NUCLEAR 1

REGULATORY COMMISSION 3Bf MARCH - APRIL 1976 E[

Oate published 8/2/76 g

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<j ISOLATION CONDENSER TUBE FAILURE With Unic No. 1 of the Millstone Nuclear Power Station operating at 100*

j i g power, the fire suppression deluge system on the =ain transfor=er in the j

i switchyard initiated without being required to do, so.

The water from this syste= apparently co=bined with salt residue on the transfor=er

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1 DI insulator causing an electrical arc-over. The are was sensed by the

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protective circuitry as an electrical fault, so the =ain generator breaker tripped open. This, in turn, caused the reactor protection

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system to initiate a trip from'100* power. The disturbance on the j-electrical system at the site also resulted in a reactor trip of the

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Unic 2 reactor from about 80~. power. Unit 2 experienced no proble=s as j

a result of the trip.

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a internal rumbling started about one minute after the trip; he presumed j

the isolation condenser had gone into service. He also noted that the E

smaller piping attached to the isolation condenser was vibrating. Other

-l personnel noted " puffs" of steam coming from the vent of the isolation condenser. One operator noticed the lights on the isolation condenser condensate return valve momentarily lose the full closed indication; there was no reason for the valve to open automatically because a reactor pressure of 1085 psig was not sustained for 15 seconds. The

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isolation condenser steam inlet valve is nor= ally open, and opening of the condensate return valve would have established flow of reactor steam through the isolation condenser.

The isolation condenser is an ll-foot diameter shell with two U-tube bundles, each with 121 stainless steel 1-inch 0.D. tubes. It is a high pressure residual heat re= oval unit used as a backup to the main con-r denser, and is designed to reject all residual heat for a period of 5 minutes after a reactor trip. The unit functions by taking steam from the reactor, condenses it on the tube side of the isolation cendenser, and returns the condensate to the reactor. With this heat exchanger, N

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residual heat re= oval and reactor cooldown can be acco=plished without use of other heat sinks such as the =ain condenser or torus.

The shell side of the condenser contains greater than 15,500 gallons of water F

which absorbs residual heat by boiling at at=ospheric pressure.

Because of the so=entary ano=aly with the indication lights, an operator was sent to check the position of the isolation condenser condensate return valve; it was found in the full closed position.

Eight minutes after the reactor trip, the main condenser was valved in to act as the pri=e heat sink. Reports from the area indicated the a=ount of steam coming from the isolation condenser vent had now dropped off considerably.

Recovery from the reactor trip proceeded s=cothly, vich the only off-nor=al condition being the continuing te=perature change in the isolation condenser. Although the te=perature change was being =enitored, it was l

presu=ed that the isolation condenser te=perature was stabiliting since the condensate return valve was known to be in the closed position.

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However, as a precaution from the continued issuance of both steam and water from the isolation condenser vent, the area was secured to minimi:e the potential from contamination. It was thought the first condensate

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return valve might be leaking, and the second in-line isolation condenser condensate return valve was closed to eliminate the boiling in the isolatien condenser. About fifteen =1suces later (65 minutes af ter the trip) an alarm was received from the iso 1.ation condenser vent radiation

=enitor. The alarm, together with reports of steam and water still coming out of the vent, caused the operator to shut the isolation con-denser steam inlet valves; steam from the isolation condenser started to decrease at this time.

After the incident, the isolation condenser shell side was en*:ered and a visual inspection revealed that one of the stainless steel tubes had failed; the hole in the tube was approx 1=ately one inch wide by two inches long. No physical damage to adjacent tubes was noted.

A review of trip sequence and plant para =eters revealed no reason for initiation of the. isolation condenser or tube failure. Pressure in the reactor never approached the isolation condenser initiation setpoint, and a review of isolation logic and relays shoved no isolation signal had been received. No thermal' shocks"had been noted.

Observations made by an operator in the i==ediate area of the isolation condenser at. the time of the reactor trip, personnel reporting " puffs" of steam from the isolation condenser vent, and a. review of the vent radiation =onitor recorder indicated the tube failure cucurred al=os i= mediately after the reactor trip.

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. The failed tube allowed reactor steam to enter the shell side of the isolation condenser for approx 1=ately 65 minutes. Actions taken during this ti=e were based on the initial evaluation that the isolation con-denser condensate return valve was leaking and that tube integrity had not been impaired.

A preliminary examination of the failed tube indicated the cracking originated on the inside surface (pri=ary waterside). The cracking was transgranular and branching, characteristic of stress-corrosion cracking. Many secondary cracks, some penetrating up to 90% of wall thickness, were found in both the bent and straight sections (near the support place) of one tube. The fractura surfaces vert covered v1th rust, and most cracks were filled with corrosion products. The crack appearance suggests these cracks did not occur during recent usage, but that they had been present in the tube for some time.

There were no evidence of chlorides either in the cradks or in adjacent corrosion deposits. One slight indication of fluoride was found. The

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cracks did contain significant amounts of silicon, calcium, and,alumytum.

None of these elements would be expected in the primary water. Their presence suggest CaSO during a past intrusion of the pri=ary cooling 4

system.

A microscopic examination of an intact tube failed to raveal any indica-tions of stress-corrosion cracking on either tube surface along the entire sample length.

After plugging the failed tube, the shall side was filled with water and leak-inspected, no additional laaks were found.

The majority of inlet ferrules, all 121 ferrules on the north and and 107 ferrules on the south end, designed to minimite thermal stresses on the tube-to-tube sheet velds, were found to have collapsed. They were to be replaced. In the outlet, only a few collapsed ferrules were e

found.

s The north side heat shield was found to have bent mounting studs aEd some misalignment and the south side heat shield mouncing studs vJre i

broken and the shield had been pulled away from the tube sheet., Heat

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shield deformation and bolt failure was attributed to thermal stress.

~3 They will be repaired. The outlet pipe thermal shield on Ihe sou'th ' side was found have a cracked weld and will be repaired.

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V Because acn-destructive eddy current exa=ination of the tubes revealed nu=erous indications ranging from 10% to 907. of through-wall thick-I.

ness, it was decided to recube the isolation condenser with 0.065-inch wall thickness Inconel 600 material.

Instru=entation changes to prevent recurrence included continuous monitoring

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J, of the shell side temperature with an alarm, a prescribed action in event of alarm actuation, and adjusting the isolation condenser vent 3

monitor setpoint close to the steady state background.

2 At the time of the incident, the potential for minor offsite contamina-I;'wasdetected.

tien was recogni:ed and technicians were dispatched to the offsite monitoring station; no measurable offsite air activity or contamination However, the tube break caused minor contamination of approx 1=stely one acre inside the fenced area. The roadways and open areas were scraped and the dirt contained for scarage. By the next day, I

all areas were clean with the exception of the inaccessible condensate storage tank = cat.

The rapid decay of the activity indicated the predem-1 inate isotopes were short lived.

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No =easurable air activity or contamination was found offsite, =ost of the released isotopes, were of short half life and no reportable radiation S

exposures occurred. The incident posed no hazard to the health and safety of the pub 11c.1,2 HPCI SYSTDi PROBLDiS Ouad-Cities 2 k' hen testing the High Frassure Coolant Inj ection (HPCI) system because of failure of the Reactor Core Isolation Cooling (RCIC) system at Unit 2 of the Quad-Cities Nuclear Power Station, it was discovered the HPCI l

system was also inoperable. The auxiliary oil pu=p kept tripping and the HPCI turbine would not start.

Because both the ITCI and RCIC were not functional, an orderly shut 4own of the reactor is required by the technical specificatien and was initiated frem 506 Mb'e.

Investigation revealed a flexible line inside the oil storage tank for the high pressure oil discharge had broken and caused the auxiliary lube oil pump to trip. The flexible line was enclosed in a wire =esh to increase its strength. The wire mesh was also broken.

The broken oil line was replaced with the sa=e type of line.

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. g The HPCI system is one of the four Energency Core Cooling Systems that provide emergency cooling water to the core over a wide spectrum of line breaks. Because the icw pressure syste=s were operable, and unit shut-down had been initiated after the RCIC system was found inoperablu, the safety implications were minimal, involving principally the availability

.I of redundant syste=s.3 Duane Arnold During a manual start of the HPCI system, operating personnel at the Duane Arnold Energy Center noted a high pressure indication in the HPCI turbine exhaust line; they manually tripped the HPCI curbine. Operating personnel had been closely sonitoring the turbine exhausc pressure because a rupture disc had blown during sis 11ar testing the previous day.

3 Subsequent investigation determined the disk of the HPCI Swing Check Valve had become separated from the hinge arm and was lying on the bottom of the check valve. The disk retaining nut and washer were missing. However, the missing nut and washer were later found in the torus.

The HPCI Stop Check Valve was disassembled in the search for the missing nut and washer. Although the nut and washer were not found in the stop check valve, it was observed that two tack welds between the valve disk and the disk retaining nut were broken.

The disk apparently separated from the swing arm in the 16-inch HPCI Swing Check Valve because, during sanufacture, the retaining nut had not been welded to the disk stud in accordance with the design drawing by Anchor Valve Company.

The apparent cause of the broken tack welds between the valve disk and the disk retaining nut in HPCI Stop Check Valve was a design deficiency.

The two eac% velds apparently were insufficient to secure the nut and disk unde

-=.tl operating conditions.

The W.he 31sk in the HPCI Swing Check Valve could have prevented pe. fD $ 3;

,I the HPCI subsystem; the disk could have lodged against the sischu 61 of the valve causing overpressurization of the HPCI tur-bine exhaust line and a subsequent trip of the HPCI turbine.

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separated from the valve sees as a result of the broken tack welds, the above analysis would also be applicable. The design function of the HPCI Subsystem is only required when nor=al makeup water to the reactor eP is not available. If operation of the HPCI system is required and the l

systes does not perform, the Auto =atic Depressurization System (ADS) in conjunction with the Low Pressure Coolant Injection (LPCI) and the Core Spray Systes are available for reactor cooldown. This event did not present a hazard to the health and safety of the public.

5 It also should be noted that the detached disk in the HPCI Swing Check l

Valve would have prevented the valve from acting as a boundarf isolation valve. However, in the as-found condition, the HPCI Stop Check Valve would have perfor=ed the isolation function and would have, isolated the

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HPCI turbine exhaust line even if the HPCI Swing Check Valve did not.

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Corrective action was to tack weld the retaining nut and disk stud in

j the HPCI Swing Check Valve in two locations in accordance with an approved y

design change. The valve disk and retaining nut in the HPCI Stop Check Valve were tack welded in four locations in accordance with vendor reco==endations and an approved plant design change.

i Sin 11ar Swing Check and Stop Check Valve installations in the RCIC were

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inspected to ensure the disk retaining nuts were adequately secured.

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The disk retaining nut in the RCIC Swing Check Valve was properly secured with a pin. Two tack welds were added to provide additional assurance the nut would remain secure. The RCIC Stop Check Valve disk retaining l

nut was found secured with two tack welds; two additional tack welds l

I were added."

3 INADVERTENT ISOLATION OF REACTOR INSTRU.ENTATION Ouad Cities-1 Unit No.1 of the Quad Cities Nuclear Power Station was at 512 M'Je while h

a routine surveillance test of the Reactor High Pressure scras switches was in progress.

'4 hen the instru=ent =echanic attenpted c3 isolate one y

of the switches prior to calibration, he discovered it was already isolated.

M3 The switch apparently had been lef t isolated after completion of the 3

previous month's testing, since no mMntenance had been perfor=ed on the switch.

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The safety significance of this occurrence was minimal; redundant pressure switches were operable. The surveillance test found the other three switches to be operational and they would have tripped within prescribed limits.

Seven days later, with the reactor at 541 MWe, a monthly surveillance test of the Emergency Core Cooling System (ECCS) pressure switches was made. One pressure switch was discovered to be isolated. Apparently, the isolation valves for the pressure switches had been inadvertently left in the closed position since the last surveillance test.

A redundant pressure switch was operable; ECCS systems would have per-formed their required function. The safety implications of both cases of inadvertent isolation were mini =al, and the public health and safety were not affected.

An additional requirement was added to the surveillance procedures to place wire security seals on safety related instrument stop valves.

This additional step should eliminate valving errors in the future.5,6 TROJAN With the Trojan Nuclear Plant at 30:: power, it was noted the steam flow indications for one of the steam lines were constant while indications for other loops were fluctuating. Investigation revealed both flow transmitters for the steam line were isolated.

2 The recorder charts indicated the sensors had been isolated for three days. During that period, reactor power varied between 0 and 30 per-cent. Apparently the steam flow indicators had not been returned to service following saintenance, and the continued operation in this period was a violation of the limiting conditions of operation.

Although automatic protection of the plant to steam line breaks was 3

reduced, adequate protection still remained. A steam line ruptura downstream of the main steam check valves would have actuated a high steam line flow signal on the other three steam lines (logic is 2 of 4) to produce a safety injection signal. It was concluded that there was no danger to the health or safety of the public.

4 To prevent recurrence, the importance of returning safety-related instrumentation to service as rapidly as possible was stressed. The operators were instructed in the importance of recording when a safety related instrument is removed from service and placing the applicable reactor protection channel in the " trip" condition.7 W

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- F PRO 3LEMS DURING FUEL MOVEMENT Crystal River N

While transferring the 55th new unirradiated fuel asse=bly fro: its shipping cent.iner to the inspection location at the Crystal River Nuclear Plant, it fell approximately five feet to the floor.

The wire cable for the fuel handling tool had pulled out of its svaged fitting.

The cable was certified to support 2,400 pounds; the fuel assembly weighed 1,550 pounds.

Some of the fuel pins were bowed, but none had ruptured. Some spacer grids were broken, the lower end fitting was bent, and some welds on the fuel handling tool were cracked.

Af ter the fuel asse=bly drop, radiation and contaminatich surveys i

indicated there had been no release of radioactive material. Therefore, the drop did not cause exposure to e=ployees or present a ha:ard to the

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public. All fuel handling slings were to be evaluated and consideration j

given to eliminating the use of svaged connections. Until this evalua-j tion is co=pleted, a different type of fuel handling to d that does not j

require the use of slings was to be e= ployed.8

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Surrv-l Unic No.1 of the Surry Power Station was in a refueling shutdown

!j following fuel =cve=ent and replace =ent of the upper internals package.

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The manipulator crane was being =oved in preparatica for latching the f

full length control rods. During move =ent, the =anipulator outer =ast

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was inadvertently driven into the upper internals package and da= aged i

the drive shaft at core location P-8.

The drive shaft was not latched I

to its control rod at the ti=e of i= pact.

The i= pact caused a bend in the drive shaf t and a minor displacement in straight edge =easure=ent of the upper guide tube.

It was concluded that the drive line was acceptable for reuse. Controlled tests indicated that guide tube =1salign=ent would not have significant effect on drive line operation.

Westinghouse Electric Corporation reviewed the occurrence and agreed that the continued use of the component would not significantly affect operation of the system.

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. As final assurance of adequacy of the drive line, an extensive red drop test program was to be conducted on this rod prior to reactor startup.

In addition, the drive line was to be inspected during the next refueling

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outage to determine if abnormal wear was occurring.

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This occurrence did not affect the health and safety of the public.9 CCNTRC' RCD PROBLEG Palisades With the Palisades Plant at 80% power, a control red dropped into the core. A turbine runback further reduced plant power to 70%, and admin-istratively plant power was further reduced to 50%. An atte=pt to retrieve the dropped red failed.

The apparent cause of the control rod drop was a shorted clutch coil.

When the clutch coil is energized, the upper and lower clutch jaws are held together and maintain rod position at the set location. When the clutch is deenergized, the lower portion of the jaw separates from the upper, and the control rod falls by gravity to a more safe position.

The shorted clutch coil did not allow the rod to be retrieved.

Pl' ant operation can continue with one inoperable control rod. However, it was decided to shut the plant down because of problems of high seal leakoff temperature in another centrol red drive mechanism.

Core flux tilts were calculated to be within the H-d ts of the technical specifications. The shorted clutch coil was replaced. There was no ha:ard to the health and safety of the public. M Robinson-2 The H. B. Roisinson Steam Electric Plant, Unit 2, had reduced power to perform weekly surveillance tests. These tests were completed success-fully and a power level increase had commenced. The Red Exercise Test was in progress. These tests check operability of the full length control rods by inserting and withdrawing them 19 steps and observing the red position indication response.

J While withdrawing shutdown Bank "A," an alarm was received on the con-trol panel. This alars prevented step movement of the entire bank of twelve control rods. A decision was made to proceed with a normal reactor shutdown as far as practicable.

With the reactor at approximately 9 power, a turbine trip occurred from a high water level in a steam generator. Shortly, the reactor was manually tripped; all rods inserted satisfactorily.

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?.i2-The alars resulted from a defective fuse in the AC power supply to a rod control cabinet.

7 The fuse was i= properly asse: bled during =anufactura, causing internittent continuity. When discovered, the fuse link had not burned apart, but was not =aking solid contact with one end of the fuse housing. This disabled one of the three phases from the AC power source 5

to the power cabinet. When rod cove =ent was atte=pted, the phase h

=enitoring card in the cabinet sensed the loss of the AC phase and initiated the alar =.

A replace =ent fuse was installed. It was believed that this was an isolated case of failure. Other fuses were not inspected because of possible link da= age from twisting.

The reactor control rods and shutdown rods were all capable of tripping throughout the duration of the occurrence, and all reactor trips were operable to provide the fullest protection possible. At no ti=a was reactor shutdo'un capability reduced by the presence of the inoperable rods.

This occurrence did not create. or threaten to create any hazard to the plant or the public.ll TAPED REACTCR S? RAY SUILDING N0ZZLES It was reported that several of the reactor building spray systes no: les at the Rancho Seco Nuclear Generating Station had, at some earlier date, been covered with tape which had not been removed.

Investigation revealed that four of the noz:les had their spray openings covered by tape, and 12 other noz:les had tape on them which did not block the spray opening and did not affect their perfor=ance. Dis-cussion with =aintenance personnel led to the conclusion that tape was placed over the noz:les to protect them during painting activity. At completion of the work, which was done while the plant was still under construction in mid-1974, the outside contractors doing the painting failed to re=ove the tape.

The four inoperative no==les represented only 2* of the 199 nozzles of the reactor spray system. Several periodic surveillance tests have been perfor=ed on both spray systa=s and the syste=s passed all require =ents of the tests. Therefore, it was concluded that the public was not endangered by the taped noz=les.12the health and safety of G

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i OCCUPATIONAL EXPOSURES 2

Indian Point Unit Number 2 of the Indian Point Station was in cold shutdown condition, having been shut down five days earlier for a refueling cutage. An individual had been assigned to determine the lighting require =ents in the general sump area beneath the reactor vessel in preparation for in-stallation of a pump.

Camma field measure =ents made several hours after shutdown and on the follcwing day showed general radiation levels in the sump area ranging from 30-150 mR/hr. However, between the time of the last field survey and the time the individual entered the area, thimbles which housed the fixed and movable in-core detectors had been withdrawn frem the reactor vessel. Withdrawal of the chimbles is required for the refuel-ing procedure and is mechanically perfor=ed at an area far removed from the reactor vessel su=p level. Unaware that the radiation field had increased considerably as a result of thimble withdrawal, the individual proceeded to the sump level. Upon reaching the sump level he checke'd his self reading pocket dosimeters (0-200 and 0-500 mrem), found them off-scale and immediately exited the area. Izmediate processing of the film badge indicated the individual had received a whole body radiation dose of 10.06 rem for the quarterly period of April 1 through June 30, 1976.

It was subsequently determined that the maximum radiation field to which this individual was exposed was approxi=ately 600 R/hr. Based on retracing the individual's steps in the identical Unit No. 3 (not yet critical), it was estimated that he spent approximately 100 seconds in the area.

The immediate corrective action to prevent recurrence included locking of the access hatch, conspicuously posting a warning sign at the entrance, partial reinsertion of the chimbles into the reactor vessel, thereby lowering the radiation levels from 600 R/hr to 50 R/hr, and placing a gamma monitor in the area to alert individuals to any increase in radia-tion fields above the postulated levels.

In view of the unusually high radiation level that can exist. in this area, a long-range investigation was initiated to determine other corrective actions for controlling personnel access which could be used in addition to controls already implemented.

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• W urg g MY ne c p-A design review of Units Nos. 2 and 3 did not find any similar situations s 37 where an operation at one location could significantly affect radiation levels at a different location. Because an unusually high radiation level can exist in this area and not be i=sediately apparent, the utility contacted all similar operating units in the U.S. and the vender to advise them of y"

this incident to prevent recurrence at another site.

g 10 CFR Part 20 requires that an individual's total accumulated exposure be limited so as not to exceed 5(N-18) where N is age of ecployee in years.

'41 thin this limit the allowable exposure in any one year is 12 For this employee, who was 32, the total per=1ssible accumulated rem.

dose was 70 rem, his actual acce=ulated dose had been 16.98 rem, including the dose received in this incident. To assure that the employee's annual allevable exposure did not exceed the 12 rem limit, he was not to be assigned to further work involving potential radiation exposure for the balance of the year.13 Zion-1 Unic Nu=ber' l of the Zion Generating Station was in its first refueling operation. When the refueling cavity was partially flooded, excessive leakage was noted from the refueling cavity into the reactor cavity.

In an attempt to discover the leakage pathway, an inspector entered the reactor cavity and, af ter remaining for 1 to 1-1/2 minutes, noted he had accumulated an exposure of 200 to 250 mrem. He wished to inspect the platform area of the annulus between the reactor vessel and the concrete shield wall, and decided he could make a rapid inspection and maintain his accu =ulated exposure under a 500 mrem limit.

He proceeded to this area, but during inspection noted the meter was pegged full scale; he is=ediately left the reacter cavity. The total elapsed ti=e on the platform was estimated to be 1 to 1-1/2 minutes.

His film badge was sent for i==ediate processing; his exposure was estimated at 8.05 rems.

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The exact level of radiation throughout the area was not known because the 58 incere detector thimbles were withdrawn from the reactor vessel into their guide tubes as required for refueling oper2cion. A subsequent survey indicated a range of at least 200 R/hr in the platform area.

5 This was the only exposure for the individus1 during the quarter.

This entry was in direct violation of approved Zion Radiation Protection Procedure. To prevent recurrence, the importance of following approved station procedures was stressed to all station personnel. To preclude l

ready access to the reactor cavity area during periods of cold shutdown, W

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administrative controls were established to have all accessess to the cavity padlocked. These padlocks were to be removed prior to proceeding to criticalit These ad=inistrative controls were to include Management verification.{4 ad HLT30LDT BAY In preparation for cleanup line repairs at the Humboldt Bay Power Plant, a foreman and worker made four trips to the lower drywell area for re= oval of flux wire guide tubes and the installation of copper

" freeze seal" cooling coils. The radiation fields in the lower drywell area ranged from 200 mR/hr to 10 R/hr.

After the foreman's second trip into the lower drywell area, his highest range pocket dos 1=eter (range 0-lR) was beyond full scale (above IR with the hairline still visible). The other individual who was performing essentially the same work and in essentially the same location, had a total indication of 1000 mrem at this time. From this measurement, a prelimin.,ry calculation indicated the highest dose the foreman could i

have received was 1200 mrem. It was decided that it would be proper for the foreman to continue the job. Both individuals made two additional trips to the lower drywell to complete the job, and each time their dosi=eters pencils indicated they had received the same doses within 50 mrem.

1 The next day the foreman was issued a new film badge and worked in a radiation area until his total estimated dose for the two days was 2550 When the film badge results were received, it was learned that mrem.

the foreman's first badge read 3500 mrem and his second film badge was a

95 mrem. The individual who had worked with the foreman in the lower drywell had a badge reading of 2300 mrem.

The most probable cause of the overexposure was the assumption that two men working side by side receive approxi=ately the same exposure and, based on chic assumptien, permitting the individual to return to work af ter his pocket dosimeter had exceeded full scale and before his film badge. result was known. A contributing factor was that the foreman was not wearing a higher range dosimeter pencil for a backup dose estimate.

To prevent recurrence, a procedure was to be initiated in which the proper actions to be taken when a pencil dosimeter goes beyond full scale would be defined, and a list of pocket dosimeter ranges to be used while working in various radiation levels would be provided. The new pro-cedure was to be reviewed with all plant radiation protection personnel and issued in the plant manual following review.

Point of

Contact:

Theodore C. Cintula Office of Management Information N

07008 and Program Control U.S. Nuclear Regulatory Commission t

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en-i nn REFERENCES

1. 2 1.

LER 76-4/10 (Supple =ent 2), Docket No. 50-245, March 24, 1976.

2.

IE Inspection Report No. 50-245/76-03, February 26, 1976.

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LER No. 75-35, Docket No. 50-265, Septe=ber 9, 1975.

3.

4 LER No. 75-23, Docket No. 50-331, April 29, 1975.

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LER No. 75-19, Docket No. 50-254, August 15, 1976, 5.

ogf 6.

LER No. 75-30, Docket No. 50-254, August 21, 1975.

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7.

LER No. 76-10, Docket No. 50-344, January 18, 1976.

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8.

LER No. 75-1. Docket No. 50-302, Nove=ber 20, 1975.

i 9.

i Special Report SR-SI-75-07, Docket No. 50-280, November 28, 1975.

10.

LER No. 75-18, Docket No. 50-255, August 28, 1975.

M 11.

LER No. 75-15, Docket No. 50-261, Septe=ber 25, 1975.

12.

LER No. 75-16, Docket No. 50-312, Nove=ber 19, 1975.

y 13.

Docket Nu=ber 50-247, April 29, 1976.

2 m_,f 14 Docket Nu=ber 50-295, April 15, 1976.

15.

Docket Nu=ber 50-133, April 23, 1976.

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