ML20024B207

Reviews Current Events - Power Reactors, Apr-May 1975. Action Required

ML20024B207
Person / Time
Site:	Crane
Issue date:	10/22/1975
From:	J. J. Barton, Toole R GENERAL PUBLIC UTILITIES CORP.
To:	GENERAL PUBLIC UTILITIES CORP.
References
TASK-*, TASK-GB GPU-2483, NUDOCS 8307070327
Download: ML20024B207 (12)

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AEC CCCU:/ENT RE'/!E'.!

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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 - Pcwer Reactors, dated: afrit _ - m4y

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, dated:

Other Review of the attached document has e.. eluded.that no action is required.

Startup & Test. Manager Date Test Superintencent Date k&MI b

%f Review of the attachte document has concluded that action is required by:

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EVENTS SELECTED FROM REPORTS SU3MITTED TO THE UNITED STATES NUCLEAR REGULATORY CC W.I5SION h

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'2" APRIL-MAY 1973 f

LOSS OF MAIN CCOLANT FUMP SEAL The H. 3. Robinson S. E. Plant Unit 2, was operating at 100". power when g

leakage fro = the first stage shaft seal of one of three =ain coolant pu=ps resulted in an alarm. The reactor power level was reduced to 36*, the y

$0 leaking pump was shut down, and the reactor tripped from a high stea=

generator flow signal. About ten =inutes later, the two remaining =ain h

coolant pumps were shut down and the co=ponent cooling water return line p

was isolated. Approximately two hours later, it became necessary to y

equalite temperature in the reactor syste=, so the pu=p with the leaking g

first stage shaf t seal was restarted. The pu=p was operated approxi=ately g

two hours, but was shut down when the second stage seal separated and the y

to the floor pu=p was now discharging a higb volu=e of pri=ary coolant F

of the contain=ent building.

Approxi=ately 135,000 gallons of pri=ary coolant water was discharged into the containment structure f ec= the leaking shaf t seals. The liquid level of the pressuriter vessel was =aintained by using three charging i

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pu=ps at high syste= pressures, and by internittent =anual actuation of g

the safety injection syste= at pressures below 1500 psig. The leakage rate through the shaf t seals reached a =axi=u= of 500 gpc. The plant

g to a cold shutdown condition by natural circulation and by vas brought the residual heat re= oval cooling =ede.

Pressure in the containment U

building reached 2 to 4 psig, but was reduced by normal ta=perature loss d

Radioactive releases were within and with the contain=ent purge syste=.

Waste water was transferred to water storage and licensee limits.

holdup facilities.

Westinghouse Electric Corporation is investigating the cause of seal g

failure, evaluating the cooldown rate of the, coolant syste= and the e

effect of flooding the lower reactor vessel.'

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h.3 IMPROPER 7ALVE LINEUP DE7 EATS REACTOR INSTRL' MENTATION

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%g During a routine surveillance test at the James A. Fit: Patrick Nuclear Power Plant, the rack isolation valves of four sets of drywell high WE.

pressure instrumentation for both safety systens were found closed.

Their associated root valves at the four drywell penetrations also were g

closed.

Because the root valves are not used in surveillance testing g

and the position of the root valves were not on a valve line-up sheet, these valves were assu=ed to be closed for the past g

seven months. A valve check of the instru=ent racks three =enths earlier had indicated y=

the rack valves were open, but the fact the root valves were closed Gg negated the design function of the dr'fwell high pressure instrumentation.

y The isolated valves were sealed open to prevent closure, and this condi-

's-j tion was added to the valve check-off lists for operations, instrument and control surveillance.

The position of the valve is to be checked prior to plant startup and during instrument surveillance testing, g

After discovery of the closed instru=ent valves, other reactor protec-

"f tion syste= instrumentation valves were inspectad, and four other valves for variable monitoring, not of safety significance, were fcund closed e

and were opened.

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Even though the high drywell pressure signal had been isolated, the I

desired functions of the safeguard systems for all accidents were not I

affected because of the availability of redundant systems, so there was no hazard to the general public as a result of this procedural 3

failure.1

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$r On returning the Edwin I. Hatch Nuclear Plant to a low power level of operation following a scram, Yarway Corporation level switches that 5

initiate core cooling systems from a low level water signal were reading.

unusually high on scale.

Investigation revealed cuo Yarways piped in 3

parallel had their equalizing valves open, causing the high upscale indications.

a The equali=ing valves were closed and proper indication

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vas restored.

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The equalizing valves apparently were not closed after the instruments 2

had been functionally tested.

I The normal startup procedure has been revised to reference valve lineup procedure, and the instrument check procedure has j

been revised to include the expected reading of these instruments tor verification of magnitude of correct approximate reading.

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i 4 Since the plant had not been to high power levels and because the improper valving was discovered at low level operation following the outage, it was concluded that the health and safety of the public was not endangered.3 J

Quad-Cities-2 With Unit 2 of the Quad-Cities Station at approximately 90% power, a

" reactor vessel low pressure" alarm was received in the control room.

A check of a redundant channel indicated reactor pressure to be corral.

Investigation determined the sensor with an alarm indication had been valved from service, and that reactor pressure was normal. "

Earlier in the day, a surveillance te'st had been complaced that required isolation of the pressure switch. With completion of surveillance, the instrument technician lef t the switch valved out.

This was the first occurrence of a switch being valved out at Quad-Cities-2.

Analysis of possible consequences led to the determination the core spray system and the residual heat removal system would have functioned properly in event of an emetgency core cooling initiation signal. Thus, there were no safety implications from this occurrence, and the health and safety of the public was not endangered."

POOR CRAFTSMANSHI? AND PERSONNEI. ERRORS Quad-Cities-2 During scheduled inspection of the twenty jet pumps at Quad-Cities Station, Unit 2, the beam bolt retainer cups and the 0.5-inch cap rcrews were found to be missing from two jet pumps. The restrainer adjusting screw on the shroud side of another jet pump was also missing. The restrainer gate keepers on five jet pumps had their welds intact, but were not fused to the restrainer gate.

The apparent cause for all deficiencies was attributed to faulty installation and workmanship. Vibrational forces on inadequately installed components could cause these components to dislodge from i

their normal positions.

f The missing beam bolt retainer cups and cap screws will not be replaced as these pieces were important only during initial assembly or removal of the pumps. The restrainer adjusting screw was installed and tack welded to its holding clamp. As a result of the faulty tack welds on W

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restrainer gate keepers, the restrainer gate bolts were 9 -

retorqued an new tack welds were placed and verified. A list of loose

}5 or =issing pa s was compiled so a search to retrieve the 1cose obj ects could be c =pl ed prior to returning the uni: to operation.

F The missing jet, p hold-down components did not lead to jet pu=p failure or to los of jet pu=p operational integrity. There were no di failures of any je pu=p hold-down components and there was no unsafe condition during pr icus periods of reac:or operation. This event did not affect the h ich and safety of the public.

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The probic=s discoverec on this inspection were si=ilar but less severe than those discovered fo lowing a jet pu=p failure of Unit 2 in August 1972. At that ti:, a jet pu=p asse=bly had becc=e dislodged ESh frem its normal position a d rotated in the vessel. Extensive inspections and repairs were performed n all Unit 2 jet pumps and they since have operated satisfac:orily.

An extensive inspection of all Unit 1 jet pu=ps during the first refueling outage in April 1974 revealed a large nu=ber of jet pump discrepancies.

These included beas bol: torque est failures, sheared restrainer gate bolts and keepers, and missing a cracked restrainer gate bol: keeper

..$M tack weld. A missing restrainer te wedge had indications of wear in E

the vicinity of the wedge, indicat g possible vertical movement of the l

wedge.

All of the jet pu=p problems that hav occurred at Quad-Cities Statica have been attributed to faulty craf e 1.

ta11ation and workmanship during the initial construction of both uni:s.

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Three Mile Island-l Three Mile Island Nuclear Station, Uni: 1, was at 997. power when a high pressure injection pu=p failed to start on signal frem the control rocm. There was a loose terminal on a Westinghouse 50-DH-P350 breaker.

The breaker was replaced and the pump operated satisfactorily.

This incident prompted further investigation into the craftsmanship and expected re11 ability of the engineered safeguards electrical circuits.

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Six of 20 motor control center uni:s (ITE Imperial Corporation Series I

9600) had wiring and/or connector deficiencies. Wire of s= aller guage g

than required by the manufac:urer's specifications, strands broken from g

multiple strand wire, poorly crimped lugs, and loosely-bolted electrical connections were discovered. The apparent cause for each deficiency was poor workmanship either during manufacture or during installation. All deficiencies were corrected.

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The connector and viring problems in the engineered safeguards motor control center did noc present a threat to che health and safety of the public. There was no evidence of overheating nor indication of immi, ant failure of a component. If failure would have occurred, operable redundant safecy' systems were available.6 c lhoun-1 a

With L it No.1 of the Fort Calhoun station in a refueling shutdown condi:1

, a high pressure safety injeccion valve failed to open on switch c d from the control room. Investigation determined the reversing neerlock of the General Electric combination breaker / starter was not m g proper contact. The reversing interlock was disassembled and inspecte a nor= ally open contact had been installed instead of a normally close contact.

Preventive mainte.ance had been performed on the breaker / starter and, evidently, the uni was improperly assembled.

The mm btenance procedure for inspection, disa sembly, cleaning, lubrication and reassembly was adequate in every res et except no functional test was perfor=ed af ter co=pletion of preventis maintenance.

The interlock was properly assembled and the valve functioned normally.

There was no danger to the. ealth and safety of the public because the reactor was in a shutdown co ition and redundant systems were available.7 Calvert Cliffs-1 Unit 1 of the Calvert Cliffs Nuc1 r Power Plant was at 100% power, and a planned discharge of the reactor colant vaste monitor tank (RCWFC) was in progress. A release permit h been obcained and the discharge was being continuously recorded on th radiation monitoring system.

Coincident with discharge of the RCWMT, e miscellaneous waste receiver tank (MWRT) was also being pumped out.

er release of the RCWMT concents, the operator discovered that the let stop valve of the RCWMT which is also an influent path from th MWRT had been left open.

Therefore, the entire contents (4000 gallons) f the MWRT had been inadvertently discharged through the RCWMT whil it was being released.

Measurements of the discharge from the radiation - nitoring system were not greater than expected values.

It was not ossible to obtain a representative RCWMT sample because the tank had b n refilled for a future planned release. Based on past analyses, a t. ical concentration W

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of the W RT is about lx105 c L uci/=1 gross beta-gn==a concentration.

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assumed concentration of this -gnitude would be corroborated by the g

monitored discharge data. The assumed activity of the release was of lictie significance with respect to the technical specifications.

To prevent recurrence, signs have been conspicuously posted in the area of the RCMiTs stating the tank influent valves =ust be shut prior to z

initiating a discharge.

has been reiterated to all operations personnel.aThe importance of adhering to written h

Oconee-2

-k During replacement of purification desinerali:er resin at Oconee Nuclear Station Unit 2, the drain valve for Unit 1 purification deminerali:er r

was inadvertently opened instead of the drain valve for Unit 2 purifica-tien deminerali:er. The control operator i==ediately identified a decreasing level indication in the letdown storage tank and monitoring of othe~r instrumentation verified tank level to be decreasing.

ge The Unit 1 purification demineralyter drain valve was shut, E.d isolated.

and the leakage

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Valves at Oconee Nuclear Station are normally identified by black C

identification tags. However, in this instance, the valve had been f

labeled with a narker.

Subsequent wetting =ade it dif ficult to distin-g guish whether the valves were for Unit 1 or Unit 2.

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the valve identifier contributed to the failure of the operator to q

adequately identify the valve to be opened.

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The prompt and timely action by the control operator detected the g

decreasing letdown storage tank level before an alars sounded.

Little, if any, resin had been transferred from the purification demineralizer.

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during the discharge and no increases in radiation level were detected.

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'41th this infor=ation, it was concluded the health and safety of the h

public was not affected by this incident.3 h

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Unic 2 of the Zion Station was operating at 857 We when a teactor trip occurred because of low level (25%) in a C steam generator coincident with i

a feed flow / steam flow mismatch.

Safety inj ection occurred 25 seconds N

af ter the trip from the high rate of steam flow together with a low-low j

Tavg (540*) steam generator temperature.

after 90 seconds when unit parameters had stabill:ed. Safety inj ection was ter=inated

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1 Prior to performing a channel calibration on steam flow loops, the operator was mistakenly instructed to select loop 522 for steam generator level control rather than loop 523. When a technician started to calibrate the feedvater regulating valve for the steam generator, the valve closed, causing the low steam generator level and feed flow / steam flow mis = acch.

The operator attempted to manually return the steam generator level to a correct value, overfed and thereby subcooled the system. The high steam flow si nals were artificially induced by the channel calibration procedure.

The inadvertent safety injection was caused by operator error in mis-reading the procedure. The personnel involved have been instructed in the correct procedure for steam generator level control calibration and to double check which channel is selected.

The safety injection system operated satisfactorily and correctly terminated upon recovery of pressurizer level. Inspection of the reactor coolant system did not reveal damage to the system and it,was concluded the health and safety of the public was not affected. M UNMONITORED RELEASE OF C0h'TAMINATED LIQUID a

Unic No.1 of the Millstone Nuclear Power Station was operating at a steady state power level of 100% while the liquid radwasta concentrator was undergoing a blowdown operation and one of the house heating boilers was being manually placed in operation. During the process of placing the boiler on line, liquid from the boiler makeup deareating tank over-flowed to the floor of the boiler room. A man leaving the boiler room found his shoes were contaminated from the water on the boiler room floor. The boiler was removed from operation and barriers were estab-lished to limit the spread of contamination.

This boiler was supplying steam to the Unit 2 heatdug system. Because of possible contamination, all Unit 2 construction workers were ordered to leave the site af ter completing personnel radiological measurements.

A radiological survey of Unit i revealed approximately 1200 square feet of floor area contaminated to a level of 80,000 dpm/100cm, 100,000 2

2 dpm/100cm in the area of the heating steam condensata surge tank, and 80,000 dpm/100c=2 in the area of the heating steam condensate i

recovery tank. Unit 2 heating steam piping was naasured at 1 mr/hr, with traps at 5 to 6 mr/hr.

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W-gk The conta=1 sated water originated with the blowdown operation of the pT~

radwaste concentrator. Heating steam to the ring sparger of the radwaste Q-concentrator had been properly valved off, but the isolation valve

?*A leaked and passed high activity concentrate to the condensate return tank. The discharge from the condensate return tank is monitored for

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conductivity and water of low concentration is returned to the boiler makeup deareating tank. High concentration discharge is diverted to the radwaste floor drain system after indication of conductivity to the y;

radwaste operator and panel alar = annunciation. ne conductivity sensor y=

was found mis-wired and, thus, per=itted high activity concentrate to 5%

enter the heater boiler makeup system without alarm indication. This instrument loop had undergone maintenance work two conths earlier and fI had not been properly checked when completed.

Spillage of conta innted water to the boiler room floor from the boiler makeup decreating tank occurred during the manual startup of the house p.:

i heating boiler. The conta=inated water on the boiler room floor flowed to an unconitored su=p that discharged to the storm drain system.

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An esti=ated 3,000 gallons of contaminated water was released un=enitored Fy to the storm drain system. The activity of water remaining on the floor I p !

was analyzed to be 1.1Sx10~2 uC1/mi gross beta. With dilution, the k

calculated average concentration at the point of release was 1.4x10~i gl uC1/ml., The average allowable daily discharge limit for Millstone is 1.0x10 uCi/ml. There were no personnel directly contaminated as a result of this occurrence. Howeveri a total of 12 pairs of work shoes were not returned to employees because of fixed contamination.

Subsequent surveys of the Unit 2 house heating piping system indicated contamination throughout the systas; all flush water and boiler drains a

were routed to the Unit i radwaste systes for processing.

The mis-wiring of the conductivity sensor was corrected and the con-ductivity instrument loop was successfully tested. Unit 1 building floor drain sump was diverted to discharga to the radwaste floor drain

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system instead of the storm drain system for decent mination of the boiler room and adjacent floor areas. The Unit 2 heating system piping vas drained, steam cleaned and refilled. Activity Javels were reduced

=h to barely above background. Local floor contamination from leakage of l [

heating system valve packing was decontaminated. Instructions were j

issued that all su=ps =ust be sa= pled and, if found to be contaminated, 3

pu= ped to the Unic 1 radwaste system. Ll W

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i CMCKS IN FEEDWATER SPARCEPJ J.

During a refueling shutdown at Unit 2 of the Quad Cities Nuclear Power Station, a liquid penetrant test on the four feedwater junction boxes revealed evidence of cracking. The test had been performed as a result of a request of General Electric Co. because of concern over cracking in several feedwater spargers installations of the same design.

There were several cracks in the heat affected zones on the piping side on each of the four spargers. After removal of the spargers, further dye penetrant examinations of the feedwater nozzle cladding detected numerous linear indications. All relevant indications were removed by grinding.

Cracking of the feedwater spargers was attributed to flow-induced vibration compounded by stresses induced by thermal gradients between the feedwater piping and reactor vessel internals. Leakage between the sparger and the feedwater nozzle also contributed significantly to vibration of the sparger assembly and imposed thermal stresses on the nozzle.

Feedwater spargers of a new design utilizing an interference fit to eliminate leakage and thus reduce vibration were to be installed prior to completion of the refueling outage.

The safety implications of this event were inconsequential because the reactor was shutdown. Although minor leakage was present, the feedwater spargers were still capable of performing their design function. There was no effect on the safe operation of the plant or to the health and safety of the public.12 Point of

Contact:

Theodore C. Cintula Office of Management Information and Program Control U. S. Nuclear Regulatory Commission i

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~y REFERENCES r

it 1.

OIE Notification of an Incident or Occurrence No. 149, May 6, 1975.

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Letter, (Niagara Mohawk Pcwer Corporation) to J. P. O'Reilly, gp USNRC, Office of Inspection and Enforcement - Region I, March 5, gj 1975. AOR No. 75-24, Docket No.30-333.

C If 3.

Telegra=s, D. P. Shannon (Georgia Power Company) to N. C. Moseley, E

"I USNRC, Office of Inspection and Enforcement - Region II, February 23, 1975. AOR No. 75-9, Docket No. 50-321.

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4.

Letter, N. J. Kalivianakis (Co=nonweal:h Edison) to USNRC, Director,

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Office of Nuclear Reac:or Regulation, May 23, 1975. AOR No. 75-16, Docket No. 50-265.

3 j

5.

Letter, N.-J. Kalivianakis (Cc==onwealth Edison) to J. F. O' Leary, y

USNRC, Office of Nuclear Reactor Regulation, February 28, 1975.

AOR No. 75-8, Docket No. 50-265.

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6.

Letter, R. C. Arnold (Metropolitan Edison Company) to USNRC, Office 5

of Nuclear Reactor Regulation, April 2, 1975. AOR No. 75-07, Docket No. 50-289.

l 7.

Letter, W. C. Jones (Omaha Public Power District) to E. M. Howard, j

USNRC, Office of Inspection and Enforcement - Region IV, March 17, g

1975. AOR No. 75-8, Docket No. 50-285.

W 8.

Letter, A. E. Lunduall, Jr. (Balti= ore Gas and Electric Company)

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s to J. P. O'Reilly, USNRC, Office of Inspection and Enforcement -

j Region I, April 17, 1975. AOR No. 75-30, Docket No. 50-317.

9.

Letter, A. C. Thies (Duke Power Company) to N. C. Moseley, USNRC, j

Office of Inspection and Enforcement - Region II, June 12, 1975.

Unusual event No. 75-6, Docket No. 50-269.

10.

Letter, J. S. Sitel (Com=onwealth' Edison) to J. G. Keppler, USNRC,.

Office of Inspection and Enforce =ent - Region III, June 12, 1975.

Docket No. 50-304 11.

Letter, W. G. Counsil (Northeast Nuclear Energy Ccmpany) to A. Gia=busso, USNRC, Division of Reactor Licensing, April 3,1975.

AOR No. 75-5, Docket No. 50-245.

12. Letter, N. J. Kalivianakis (Coe=onwealth Edison) to J. F. O' Leary, USNRC, Office of Nuclear Reactor Regulation, March 21, 1975.

Docket No. 50-265.

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