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,f,-. NRC 00CLWEt;. REVIEW-

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Blant/ Unit - //-- ',*

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The attached fiRC docume#*ns has been reviewed for test program and designi,

:=cdification requirements for the above Plant /Ucit.i

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CCCUMENT: Operating Experience, dated:,,

f Current Events - Powr Reactors, dated: |- & /|/ff).

, Other , dated:

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Review of the attached docment has concluded that no actica is required.m.

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8 1 -/S~-78Stay & i'est Manager Date

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est superintancent . Data.

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Review of the attached document has concluded that action is required by:.

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Problem Report (s) , , , e___

has/have been issued. /gmer

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| Startup & Test Manager Date1

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Test Super 1ntencentDate

STRIBUTI0tt: R.W. Heward, Jr. W.T. Gunn E.D. McDevitti % J.E. Kunkel M.A. 4elson R.J. Toole

J.T. Faulkner File.

8307110034 780115DR ADOCK 05000289

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UNITED STATESCURRENTEVENTS

NUCLEAR

REGULATORY.

POWER REACTORS COMMISSION

THIS COMPILATION OF SELECTED EVENTS IS PREPARED TO DISSEMINATEINFORMATION ON OPERATING EXPERIENCE AT NUCLEAR POWER Pt. ANTS IN A

-(" $ TIMELY MANNER AND AS OF A FIXED DATE. THESE EVENTS ARE SELECTED FRCMPUBLIC INFORMATION SOURCES.. NRC HAS, OR IS TAKING CONTINUCUS ACTIONON THESE ISSUES AS APPLICABLE, FROM AN INS?ECTION AND ENFORCEMENT,LICENSING AND GENERIC REVIEW STANOPOINT..,.

; 1 SEFlPIER - 31 Ctiutx 197i

(PU8LISHED DECEMBER 1977)

OPERATOR ERROR

on January 11, 1977 while the Fort Calhoun Station Unit I wasoperating, water from the Refueling Water Storage Tank was pumped

.into the containment through the containment spray header due to anoperator error. '

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During the performanca of aand containment spray pumps,quartarly test of the safety injectionthe operator noticed an increase inthe containment sump level,approximately tan minutes after thelow pressure safety injection pump had been startavi.3300 gallons of water had been pumped to the contatnment.ApproximatelyAbout one..;;ygg minuta later the ventilation isolation actuation signal was received. .

At this time the operator realized he had failed to follow the sur-veillanca procedures and had left the discharge valve of the icw headsafety injection pump open. He immediately secured the pump...

The Reactor Coolant System was checked for leakage and containmententry was made approximately one hour later. Inspection revealedthat a discharge from the containment spray nozzles had occurred.A few minutas later power reduction was startad. A second containmententry was made about an hour later, after containment air samplesconfirmed that a full face mask would provide adequate respiratoryprotaction for'the levels of radioactivity in the building. A ,

M detailed inspection revealed no serious deficiencies and no electrical?] 2 grounds; the power redue:1on was terminated at a power level of 83t.'-

Although the operator had not followed the procedure and the dischargevalve was open, the containment spray header isolation valve (HCV-345)

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and the low pressure safety incross-connect valve (HCV-335) jection to containment spray header

should have prevented the event. Theelectric / pneumatic ccnverter on HCV-345 had failed and both red andpartially open. green position indication lights are on, indicating the valve was

Operster had takan local control of the valve in an attempt toPrior to the event the auxiliary Building- Equipment"

completely close the valve.After about 1/2 inch of stem travel, the

position as demanded by the valve positioner. operator removed the valve pin and the valve went back to its previousThe third valve ~ (HC/-335)

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but no cs...ctive action had been taken.in the incident had a leakage problem that had been previously identified.

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The pneumatic relay on valve HCV-345 was replaced and valve HC/-335.w

repaired.Valve HCV-344 and HC/-345 are now required to be placed

in the test mode prior to operating the icw pressure safety injectionpump or contain spray pumn for testing. This mode along with verifi-<

cation of an annunciator will ensure that both of thage valves.ars-

in the fully closed position prior to pump operation. '

VALVE MALFUNCTIONS,

1. PMaary System DepressuMzatier.,

On September 24, 1977, Davis Bessa Nuclear Power Station Unit~

No.1 experienced a depressurization when a pressurizar powerrelief valve failed in the open position., The Reactor CoolantSystem (RCS) pressure was reduced from 2255 psig to 875 psig inapproximately twenty-one (21) minutas. At the beginning ofthis event, steam was being bypassed to the condenser and the

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reactor themal power was at 253 MW, or 9.5%. ElectricityM was not being generated. Tha following systems malfunctionedduM ng the transient:.

Steam and Feedwater Rupture Control System (SFRCS).a.

b.. Pressurizar Pilot Actuated Relief Yalve.

No. 2 Steam Genentor Auxiliary Feed Puso Turbine Governor.c.

The event was initiated at 2134 hours, when a spumous " half-trip"occurred in the SFRCS, resulting in closure of the No. 2 Feedwatert

Startup Valve and loss of flow to No. 2 Steam Generator. Approxi-W,mately one minute later, low level in the No'. 2 Steam Generaterp caused a full SFRCS tM p, closing the Main Steam Isolation Valves

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(MSIV). The loss of heat sink for the reactor caused the RCS ltamperature, pressure, and pressurizer level to rise.'

The RCS pressure increased to the pilot actuated relief valvesetpoint (2255 psig) and the valve cycled open and closed nine

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Meanwhile, the reactor operator observed the pressurizar leveltimes in rapid succession, failing to close on the tanta opening.-

increase and manually tripped the reactor about one minuta afterMSIV closure (two minutes into the transient).At this pointthe RCS pressure was approximately 2000 psig and decreasingwhile the pressurizer level had reached its maximum initialrise of about 310 inches. The RC3 pressure continued to decreasedue to the open relief valve and upon reaching 1620 psig approxi-

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mately three minutes into the transient, actuated Safety Features. including high pressure (watar) injection and containment isolation.

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Approximately five minutes into the transient the rupture disc', '

on the pressuM zer quench tank, which was receiving the RCSblowdown, burst..

by the actuation of containment isolation, which had 1solatedBursting of the rupture disc was aggravatedthe quench tank cooling system, resulting in expedited pressuM-zation of the quench tank.

The RCS continued to blow down through the open pressurizer-.

power relief valve and the quench tank rupture disc openinguntil pMaary coolant saturation pressure was reached, aboutsix minutas into the transient. The femation of steam inthe RCS caused an insurge of water into the pressurizar. Thisinsurge and the high. pressure watar injection then restoredpressuMzar level to about 310 inches after nine minutas intog the trsnsies.t.

Approximately thirteen minutes into the transient, the secondaryside of the No. 2 Staam Generator went dry. About fourteenminutes into the transient,. the operators noticed the low levelcondition and found that the auxiliary feed pump was operatingat reduced speed. Manual control of the auxiliary feed pumpwas started and water level restored to the No. 2 Steam Generator. .

At approximately 27 minutas into the transient, the operatorsj discovered that the pressurizar power relief valve was stuckopen. Blowdown via this valve was stopped by closing the block'

valve, thus taminating the reactor vessel depressurization. The| RCS pressure recovered to normal and cooldown of the systam followed.!'.

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The reason for the spurious " half-trip" of the SFRCS has not yetbeen detemined. An extensive investigation revealed severalloose connections at taminal boards, but nothing conclusive.

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Investigation into the failure of the pressurizar pilot actuatedrelief valve revealed that a "close" relay was missing frem thecontrol circuit. This missing relay would nomally provide a." seal-in" circuit which would hold the valve open until thepressure dropped to 2205 psig. Without the relay the power relief

l valve cycled open and closed each time the pressure of the RCSwent.above or below 2255 psig. The rapid cycling of the valvecaused a failure of the pilot valve stas, and this failure caused,

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the power relief valve to remain open. -

It was detarmined that the auxiliary feed pump did not go to fullspeed because of " binding" in the turbine governor.

The transient was analyzed by the ff533 vendor and datamined to bewithin the design parameters analyzed for a rapid depressurization.

With exception of the above noted malfunctions, the plant functionedas designed and there was no threat to the health and safety of thegeneral public. ,3 ,

2. Feedwater Isolation Valves

On two occasions in July, at tin Tr'ojan nuclear plant, a hydraulic'

feedwater isolation valve failed to close upon receipt of a closesignal. All other equipment required to operate, functionednomally. '

The first failure Jul.y 6,1977, had been attributed to animproperly assembled solenoid in the hydraulic actuator.Investigation of the second failure indicated that both eventswere due to a lack of sufficient hydraulic pressure.

FaiTure of the valve to close was caused by the pressure regulatorleaking and failing to close down to regulata the pressure. Thiscaused the hydraulic system on the valve to be drained down toa point that the valve would not operata. Inspection of theregulator revealed that a locking screw on the regulator adjusting:

knob was loose and would allow the knob to vibrata to any position.s With the regulator improperly set it would not close down to.

I regulata pressure and would allow the hydraulic fluid to drainbefore the hydraulic operator could

. as discovered on two other valves, function. A similar problesw although the maladjustmentwas not sufficient to prevent these valves frem operating.

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All of the regulatcrs were reset and the adjusting knobs werelocked in place so that they could not vibrata loose. Thsisolation vaadjustments.{ves were tasted satisfactorily following these i-

3. Off-Gas System Valves,

At the Oystar. Creek nuclear generating station on August 27,1977, the reactor building ventilation system isolated and thestandby gas treatment systam (SGTS) automatically initiated.

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Investigation revealed that at approximately 1850 hours a station*

employee perfonning housekeeping duties in the main control roomaccidently caused the augmented off gas (ACG) itd4 switch to movefrom " isolate and bypass" to the "isolata" position. This resultadin the off gas valve and the off gas drain valve going closed,and since the A0G was not in service the gas flow was stopped. Theisolation of the reactor building ventilation system and initiation'

of the SGT3 occurred at 1905. The two off gas valves were openedfour minutes latar and the SGT3 was secured. The reactor buildingventilation system was returned to normal at 2000 hours.;

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] The off gas drain valve did not seat properly and was not leaktight. This condition allowed the gaseous radioactivity within,

the isolated off gas system pioing to travel *up through the stacksump in the stack base and fill the air space in the ventilation-

tunnel.( When the radiation level in the reactor building| ventilation duct reached a level of 17 mr/hr the monitors located

next to this duct initiated the SGT3.

The safety concern Associated with this event is the possibility:; of a submergence dose a person would have receivef from the radio-active gaseous atmosphers if they were in the tunnel area. Theatmosphere in the tunnel area is processed through the radwastaventilation systam, which contains both roughing and absolutafilters, prior to axhausting through to the stack which is '

mnitored. The maximum radiation level sensed in the tunnel was*

26 mr/hr.

No personnel exposures or reiseses to the environment resultedfrom this event. The licar.see is investigating the feasibility ,

of installing an alans to alert operations personnel to the closureof the off gas valve when the A0G is out-of-servica.5

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SPALL PIPE BREAX ANALYSIS

On June 9,1977, an orderly shutdown of the Yankee Nuclear PcwerStation (Yankee Rowe), a pressurized water reactor, was initiated

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by the licensee because of an error discovered in the Emergency Core'

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Cooling System (ECCS) perfomance analysis.

Yankee Atomic Electric Company (YAEC), the licensee, notified the \

Nuclear Regulatory Ccmaission (NRC) that an error had been discoveredin a particular small break loss of coolant accidentwhich pemitted reactor operation with Core XII in a m(anner) analysis,LOCA

conservative than assumed in the original analysis. less. .-

While performing a review of the analyzed small break accidents fort

the Core XIII reload, the YAEC Safety Analysis Group detemined onJune 7,1977 that an incorrect fluid flow resistance calculationwas made in the safety injection Ifne break analysis.characteristics study had taken credit for the 2-1/4 inch safetyThe fluid flow

a tank which supplies borated water to the reactor core in the eventinjection line thermal sleeve to retard spillage from the accumulator-

of a reactor coolant system pipe break.sleeve should not have been includM in the flow calculation, as aThe flow resistanca of thenew worst case pipe break was identified in a 4-inch diameter linesection.

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'The recomputed decreased flow resistance allowed increased accumulatorflow to be caTeislated for the breakreflood capability of the ECC3. pressure to less than had been assum,ed, thus decreasing the coreand decreased the ECCS supply

This corrected flow resistanceassumption was used for the accident analysis of the present core,Core XII, which was operating at 79% of rated power in a coastdownprogram prior to the June 9,1.977 shutdown. Operation of the reactorwith Core XII cemenced in December 1975.

Upon discovering the error, the licensee reduced power level to 300megawatts themal (50% rated power), which was believed to conservativelyaccannodate the analysis error.

During subsequent analysis, however,the licensee was unable to assure himself that the 10 CFR 50.46 limitson peak fuel cladding temperature could be mainta'ned for the postulatedsmall break.

Therefore, the faciltty was shutdown pending resolution,

had been previously scheduled to cannonce on July 2,1977.of this matter and to proceed with the Core XIII refueling outage which

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The licensee subsequently perfomed an approximata best estimataanalysis of the postulated worst case small pipe break, which includedCore XII operation. assumptions based on actual facility equipment availability duMng

The results of this analysis indicated that the .

calculated peak fuel cladding temperature was well below 10 CR 50.46limits.The more conservative 10 CFR 50 Appendix X reanalysis of Core

been exceeded in the event that the safety injection pipe break hadXII operation, however, indicated that 10 CFR 50.46 limits might haveactually occurred.

prior to returning theCore XIII the licansas: plant to operation after refueling of

1) performed flow measurements tastsjto determine the actual flow resistance through the safety injection

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piping; 2) changed the flow resistanca in the safety injection lines,,

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by an ECCS modification; and 3) analyzed appropMata pipe break.

iaccidents in accordanca with 10 CFR 50 Appendix X cMtaria.i Thechanges and results of tests and analysis were submitted to the NRCand were approved prior to restart of the plant after the refueling.6-7{

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DIESEL GENERATOR TRIP'

ij During a loss-ofwe test on August 2'6, 1977,*the E-4 diesel of

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the Peach Bottom Atomic Power Station Unit 2 started properly as a:i

result of the undervaltage condition,'but tripped irmediately. ThistMp was caused by the overspeed mechanism.,

The circuitry was reset,an adjustment was made to the mechanical governor to ifmit the dieselspeed duMng a start and the unit was started successfully. Because-the exact cause of tha tMp was not firmly established, surveillance

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tasting of the diesel was increased from once a week to once per shift.|

During one of these tests, on August 27,1977 . the diesel trippedagain.Another adjustment was made to the mechanical governor, the

load capability was checked and several successful starts were perfomed.|Onca per shift surveillanca was continued.

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On August 29, 1977, the diesel again tiMpped on overspeed and wasdeclared inoperable. The diesel was then operated in excess ofsynchronous speed in order to detarmine the exact speed at which theoverspeed mechanism would function. This test determined that thej diesel would trip at 940 rpm instead of the desired setpoint of 990

i The tMp mechanism was adjusted to 985 rpm by,a manufacturer'si rpm.

representative and diesel was started twice, successfully.,

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Investigation into the cause of the change in the trip settingydatarmined that during the diesel maintenance in June 1977 a camshaft:

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was replaced.In order to replace this camshaft the overspeed mechanismhad been removed. When the overspeed mechanism was replaced, semenecessary shims were not installed. Although this was the only diesel

requiring this maintenance during the annual check, the other dieselswere operated up to a speed of 945 rpm to verify proper operation.None of these diesels tripped cn overspeed. .

Analysis of this event revealed that a deficiency exists in themaintenance procedure associated with the diesel yearly inspectionand the post-maintenance testing procedure. These procedures willbe revised. to correct the deficiencies.8 - -

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ELECTRICAL FAULT,

Cn July 13, 1977 while the personnel at James A. Fitzpatrick nuclearpower plant were conducting refueling operations a short in a cablecaused 600 volts AC to be introduced into a 115 volt circuit.The600 volt AC supply for the refueling bridge and the 115 volt ACcircuit for refueling interlocks are both located in the same cable.Flexing Of the cable with bridge motion over the core caused the cableto short internally.

The introduction of the 600 volts into the 115volt circuit caused ninetewn relays in tNE rod manu'al control systaato burn out . All of the refueling operations were haltad until theinterlocks were repaired. The rod worth minimizer and rod sequence,

control sysems were also checked for damage.,

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.A modification is being prepared that will remove the 1T5 volt AC.

!. ?interlock circuit from the ca

| This will prevent recurrence.gle carrying the 600 volt AC supply.'

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PIPE CRACK I-The Brunswick Steam Electric plant Unit 2 was in het shutdown and 1preparations were underway to startup the unit when the Shift Foreman y)noticed a small leak of the recirculation loop suction piping. This 'l

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discovery was made during the closecut inspection of the drywil.Q,

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Investigation revealed the leek was from a crack in the socket weld :=|

b}l on a three-quarter inch test connection 900 elbow that was nonisolable ;

and the plant was placed in the cold shutdown condition. The cracked 7I

pipe was cut out of the system and the connection was capped. Similar *

connections on both Units 1 and 2 were dye-penetrant checked with no.

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other indications of cracks. ~

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weld metal and intergranular stress corrosion in the heat affectedNurther investigation revealed that the crack was contained in the.

zone of the base metal was ruled out.the internal and external diameters of this section of pipe reveal dA dye-penetrant inspection ofno other cracks.weld joints showed that a proper gap was present between the socketThe inspection of the internal diameter of the socket

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I and the pipe end.!

Based on a stress analysis and the observed condition of pemanent.

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deformation of the failed area, along with the location of the crackit is concluded that the initial crack was caused by stress concentration-

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in the weld fillet area.,

result of workmen (during constructionIt is believed that this deformation was theii

This use of the pfpe for this purpose p)lus vibrational stress resultedusing the pipe as a step.j in the failure.l

A visuai inspection of similar piping on the other loop of Unit 2jand both loops of Unit I revealed no deferination as was observedon the failed pipe.,

remaining pipes is such that they are not. likely to be used as a steoIt was also noted that the location of the three| |

3or support because of physical interferences.ibe supported to protect them from experiencing excessThese three pipes will;

!leading and vibration, or will be removed and capped.gegternal

i Point of Contact:1 Joseph I. McMillen:

Office of Management Infonnation[' and Program Control

' U.S. Nuclear Regulatory Consission4

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_ REFERENCES

1.LS 77-2, Occket No. 50-285, January 31, 1977.

2.LER 77-16, Occket No. 50-346, Octcher 7,1977. \

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3.Supplement to LS 77-16, 'Occket No. 50-346, November14, 1977.

4. LS 77-23, Occket No. 50-344, July 29,1977.,

5. LS 77-21, Occket No. 50-219, September- 23, 1977.j 6. LS 77-30, Occket No. 50-29, August 3,1977..

! 7.,

Sumary of June 17 Meeting, NRC-YAEC, June 22,1977.8.

LER 77-37A Occket No. 50-277, September 9,1977.9. LS 77-43, Docket N3. 50-333, August 11, 1977.

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; 10.LS 77-7, Oceket No. 50-324, February 28, 1977.'

j 11.Supplanent to LS 77-7, Occket No. 50-324 Septamber

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30, 1977.!

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