ML19322D923

Licensee Event Followup,Procedure 92700B.NRC Aug-Sept 1975 Current Events Encl

ML19322D923
Person / Time
Issue date:	01/01/1979
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
Shared Package
ML19322D920	List: ML19322D919 ML19322D921 ML19322D922 ML19322D923 ML19322D924 ML19322D925 ML19322D926 ML19322D927 ML19322D928 ML19322D929 ML19322D930 ML19322D931 ML19322D932 ML19322D933 ML19322D934 ML19322D935
References
PROC-92700B, NUDOCS 8003110204
Download: ML19322D923 (20)

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Licensee Event Follovuo Procedure No.: '92700B r

Issue Date:

10/1/76 E

P' SECTION I D;SPECTION OEJECTIVE Ascer:ain whether the licensee's review, corrective action, and report of. the identified event and associated conditions are adequate and in confor=ance with regulatory requirements, Technical-Specifications, license conditions, and licensee procedures t.nd controls.

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Licensee Event Followup j'

Procedure No.: 92700B k;

Issue Date:

1/1/79 m.

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INSPECTION REQUIREMENTS E

E E-For events selected for followup, conduct record review, direct i

observation, or discussion with licensee personnel to the extent E6-necessary to complete the applicable inspection requirements.

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1.

Corrective Action - Technical 1.-

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a.

Ascertain that corrective action is appropriate to 5;.=j correct the cause of the event.

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b.

Verify that corrective action has been taken.

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c.

For corrective action not yet complete, verify that C

responsibility i.es been assigned for assuring comple-l t

tion thereof.

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d.

Verify that generic implicatioris if identified were

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e.

Detennine whether corrective action taken or to be taken j

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is adequate, particularly to prevent recurrence.

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2.

Safety of Operations - Technical t

a.

Ascertain whether the event involved operation of the h-facility in a manner which constituted an unreviewed i

E safety question as defined in 10 CFR 50.59(a)(2); or, l

l-7 for facilities or operations not covered under 10 CFR 50, L..

in such a ranner as to represent an unusual hazard to j..

health and safety of the public or environment.

b.

Ascertain Mether the event involved continued operations' i

If in violation of regulatory requirements or license i..

f conditions.

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'eporting Requirements - Adninistrative i

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Ascertain that reporting requirements have been met by verifying that:

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me The report accurately describes the actual event.

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b.

The safety significance s'.ated in the report is W

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consistent with details of the actual event detennined

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Licensee Event Followuo

J." i Procedure No. : 927003

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Issue Date:

1/1/79

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c.

The reported cause is accurate and the LER form, if required, reflects the proper cause code, r~

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The report. satisfies the reporting requirement with

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respect to infomation provided and timing of submittal.

4.

Licensee Review - Administrative a.

Verify that the event was reviewed and evaluated as

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required by approved procedurc: and administrative controls.

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b.

Verify that personnel within the licensee organization l

were notified of the event as required by Technical i

Specifications license conditionb -or -appr-ove4 l

procedures.

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Verify that review and evaluation of the event included l

assessment of generic implications.

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Verify that review and evaluation of the event included assessmnt of personnel error and procedural adequacy.

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Verify that the event was reviewed to detemine whether it is a recurrence of past events. 't r

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If the information reported to the NRC is found to be significantly in error, verify that the licensee submits a corrected report.

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6.

Document closeout of event followup in the inspection report.

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Licensee Event Followuo

$5 Procedure No.: 927008 K

Issue Date:

1/1/79 g;;

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SECTIC'i III A_.

E IN5?ECTIO'l GUIDANCE E'

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General guidance on the structure of reporting requirements is is provided in MC 907123, Sectica III, item 1.

The depth of onsite jEf followup is based on the safety significance, complexity of tech-EE-

.nical probleas and generic ir.nlications of the event.

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judgement with concurrence of Regional canagetent should determine

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the extent of followup needed on each report.

It is neither i

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Ei; required nor desired that all reports be folicwed up onsite.

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The inspection requirements are divided into technical and admini-strative categories which serve different punoses.

The technical ET aspects of the event include operational details, cause, corrective 8

gl action, prevention of recurrence and generic implications.

Admini-Mi strative aspects address the licensee's management system for EF processing reports. Within each category the inspection requirements g.)

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are necessarily flexible to the extent that the type of event may cause som individual "line items to be inapplicable.

It is not Ep

. necessary that such re-quiremEnts be completed.

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Followup entails onsite verification that couldn't be done during

-55 53 2 in-office review, and is in addition to, not duplicative of,

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.b. Corrective action items of a icnc ter:a nature, such e

as a design change, should be tracked to completion

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by the inspector.

c.

Fortal requirements should be established in the

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licensees' adr:inistrative cor. trol program to assure l

that corrective actions have been completed.

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d.

The applicability to, or susceptibility of, other systems or components in the affecte:! unit and other units at the site, is the extent to which generic-implications should be included in corrective action.

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Corrective action should generally include action taken b

at the tir:e of the event to eliminate the cause or to L&:

mitigate consequences, action taken to correct the specific fault cr failure (maintenance, repair, replace-M:.:

ment, procedure change, special administrative control, iE-EE-etc.), and action taken to reduce the probability of, W1

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Licensee Event Followup gg Procedure No.: 927008

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Issue Date:

1/1/79 0

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or.to prevent recurrence (design change, operator Ef5 retraining, maintenance practice, work controls,

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These aspects of corrective action apply Es in varying degrees to a specific event, and as such, gi inspector. Judgement is necessarf in this determination.

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a.

This item directly supports the IE responsibility to i.F determine whether licensed operations are being conducted EK safely. 10 CFR 50~.59(a)(2) should be used as guidance in assessing safety of operations of facilities not licensed g2 under 10 CFR 50.

Even though 10 CFR 50.59 is not a require-EE:

ment for these facilities or operations, it delineates the types of items that should be considered in assessing l54 Ei; i

safety of such licensed activities.

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Examples include safety limits, limiting safety system E

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settings, limiting conditions for operation, limiting control settings and other regulatory requirements asso-En ciated with the event. Inspection should cover items M"'

such as required settings were not exceeded and redundant

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systems required by Technical Specifications or license Es.

conditions were operable.

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In determining that the rep:rt is accurate, items such as l

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s the following exacples that are associated with the event

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and applicable to the repor'ing' requirement should be verified:

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!iature and extent of radiation exposure to employees

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l llature and extent of radioactive releases.

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- Extent to which an ir.strenent was found out of E

a calibration or outside tolerance allowed by i

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Technical Specifications or license conditions.

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Refer to the reporting require:nent to determine acceptability.

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MC-Interpretations, R.G.1.16 aalifies degraded mode and

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report type as to prompt notification or 30-day timing.

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Approved procedures should assign responsibility to Emr personnel for prompt review, evaluation, determination 1

EE of the cause, and notification of licensee personnel of 5.=

the event. Safety cocmittee reviews of reportable E;

occurrences will be verified on a sampling basis during El.

the perfornnce of 407003.

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gy J.3 ensee Event Followuo hi:

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Procedure No.:

927008 EE:

.zr Issue Date: 1/1/79

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Results of these assessments should be reflected in 5

'detemination of cause and corrective action for E:l-events in which these issues are germane.

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Licensee adninistra'tive procedures should be adeaune 4.

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to identify whether an event is first of a kinc. or of

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a recurring nature.

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The licensee should submit a corrected report to the NRC if (narrative and/or LER data fom, as appropriate).

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is detemined-that the incorrect information was purposely

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reported or if the licensee frequently r.eports misinfonna-15E-1 tion, enforcement action would be appropriate.

The threshold Q[

of significance of errors including omission, above which a E:l

, corrected report is-required, involves inspector judgement, E!..

and should be connensurate with consideration of citing the E?

licensee for failure to report, which infers failure to is=

report accurately.

Errors of lesser significance should E:?i be discussed with the licensee' with the intent that future Eu reports be correct, but insistence on submittal of a E&

corrected report for the specific event is not warranted.

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Documentation of findings under this module in the inspection

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in a traceable manner.

If nore than one inspection is

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Reference 3 6l h1 m

c5 UNITED STATES

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4 CURRENT EVENTS NUCLEAR REGULATORY s

POWER REACTORS >

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EVEh'ES SELECIED FROM REPCRTS SU3MITTED TO THE UNITED ' STATES NUCLEAR

[. I REGULATORY COMMISSION F'

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1. n AUGUST - SEP'TEMBER 1975 9:'

(i.n C-5 CRACRS DISCOVIEED IN COLLET EOUSINGS OF CONTROL ROD DRIVES During a refueling outage at Unit 3 of the Dresden Nuclear Power Station

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and sile overhauling a control rod drive, a crack was discovered in the h

collet housing short tube.

Four other control rod drives were available P#

for scrutiny; inspection revealed each of their collet housings to be cracked.

In each case, the cracks occurred in the' collet houcing short U

tube below the water ports in the a ea of increased wall thichaess.

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(O Subsequent inspection of eighteen control rod drive nechanisns revealed h

that eleven rods displayed sene indications' of cracking in the collet

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housing area.

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g General Electric Conpany's Nuclear Energy Division was advised of this I

possible generic problem.

Ern-N tion of'their test control rod drive

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nechwisns revealed cracks of the collet housings nearly identical to the t

four control rod drives ex=-",ed at Dresden-3.

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e General Electric had been zware cf similar cracking on test collet housin5s of control rods that had been scran-cycled 2000 times, and nore severe.

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cracking on nechanisns scran cycled more than 4000 times.

However, there were r.o indicatiens that cracking would develop within the expected F

lifetime of 200 scrans fer the control rods at Dresden-3.

The collets with cracked hcusing were replaced with new assenblies.

f1 Future actions will be deternined by the outcome of tests now in progress.

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At present, Gene al Electric and Argonne National Laboratory are conducting k;

independent metallcrgical tests to determine the cause of cracking.

f Although ir has not been substantiated, the cause of cracking may be related to the te perature cycle a centrol rod drive experiences during a reactor

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L The 2000 and 4000 cycle scran tests performed at G.E. demonstrated the k

probability of total collet housing failure to be quite remote.

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collet housing does not function as a pressure barrier and is subject to j

stress vastly less than the yield strength of collet housing metal.

In a supposed possible worst conditions accident, a number of collet housings T l failing si_.ultaneously, localized core damage could result from abnormal rod patterns and power levels. However, even in this unlikely event, a

.y standby liquid centrol systen would be available to reduce reactivity and j

maintain the reactor in' a shutdown condition; all radioactivity would be j

contained within the reactor vessel or the standby gas treatment systen; and there would be no danger to plant personnel or the public. I i L 3 e t '.

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STE.Of GDERATOR TU3E LEAK G.

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At Unit 2 of the Point Beach Nuclear Plant, operating personnel noted an l

upward trend on the air ejector radiation and blowdown monitors, indicative i u of pri=ary-to-secondary steam generator leakage. The primary-to-secondary h

e leak rate was calculated to be 0.23 gpo, a rate near the normal average of 0.2 gps, but operating personnel began securing systems in anticipation of l i a blowdown / shutdown. Five and one-half hours later, the primary-to-secondary 1eak rate had increased to 0.4 gps, and four and a half hours af ter that, g

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the blowdown monitor alarm was received.

Thirteen and one-half hours later, an orderly shutdown of Unit 2 commenced at a rate of approximately 100.W/ hour.

Subsequent eddy current inspection identified a f ailed tube in the "5" steam generater.

The failed tube was on the periphery of the tube bundle slightly above the top of the tube sheet, and the appearance of h

a relatively clean cut and roughly circular hole' indicated a manufacturing f

defect or the result of da= age.following manufacture.

The f ailure appeared y

to be randon in nature and not connected with previous gerieric proble=s of M

wastage in the kidney shaped high heat flux zone of the hot leg.

1g Two tubes, in addition to the leaking tube, were discovered to have signif f cant defects. One, a 44% defect, was located at the third support.

A second, o

with a 58% defect, was located approrhtely one-inch above the sixth support. These tubes had previously been measured with a 20 to 30% defect, and a 40% defect respectively.

The failed tube had never been examhed in service.

1 Of the 712 tubes tested during the eddy current program,150 appeared to a.xhibit a loss of ovality of 0.002-inch or greater.

The steam generator t

reufacturer has advised that tube vibrat!on from crossflow of water on a tube may be a contributing factor to the :'ss of ovality.

There was no mea.surable metal loss with loss of ovality.

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The switch to all volatile water chemistry treatment at both Units 1 and 2 appears to have inhibited the tube vastage problem previously discovered in the ke y-shaped high heat zone of the steam generator hot legs. No e

new indications of wascage in this heat. none were observed during the inspection.

In addition, although sludge depths of up to four inches were measured on the tubesheet by eddy current exeination, this sludge appeared to be haraonious with the tubing. Sludge lancing, therefore, was not perfo ned during this outage.

L A secondary-to--pri=ary 800 psig leak test was perforned with satisfactory

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results.2 7

FAILUEE OF SATITY RELIEF VALVE

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With a reactor pc er of approxi=ately 10% at Unit No. 2 of the Brunswick Steam Electric Plant, the "B" safety relief valve inadvertently opened.

An attecpt to close the relief valve by placing the control switch to close failed.

(A violation of the emergency instructions occurred when

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the reactor was not =anually shut down when it was deter =ined that the pO relief valve was stuck in the open position.)

Concurrent with attempt to seat.the relief valve, an attempt was nade to initiate torus cooling with one of the residual heat removal loops, but the service water supply-valver to the heat exchacger failed to open. A redundant loop vas 4-ediately placed in the torus cooling mode.

When the decision was made to shut down the reactor, the High Pressure Coolant Injection (E?CI) operated for only a limited time because of high torus level. c*rinen. it.was. apparent.that manual operation;of the BPCI-could not supply clean water to the reactor, the main steam isolation valves a

were closed. This action resulted in a reactor scram.

Reactor pressure i

decreased rapidly, and continued to decrease until the pressure reduction was stopped at 72 psi by apparent seating of the relief valve.

There was no damage to the torus structure or relief valve discharge. pipes, and inspection indicated all components reacted normally to the discharge failure. A specific cause for the blevdova incident was not discovered.

All relief valves were actuated successfully at 50 psi during subsequent reacter heatup, and all relief valves net capacity checks successfully.

No prchlen with valve operation was identified.

Seven days later, v'th the reactor at 8-9% power, and at 600 psi pressure,

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it was observed that 'the te=perature of the discharge of the same relief s

valve was abner--17 high. The relief valve was cycled three times he "l

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the open relief valve, but the valve still did not reseatAdjacent valves were

-d sure was at 475 psi and decreasing so the reactor was nanually scra=nedReactor pres-N.7 i

During the blevdevn, several atte: pts were =ade to reseat the relief d

valve:

cace at 184 psi, once at 82 psi, and once at

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half honrs af ter the relief valve inadvertently opened, the valve appeared 49 psi.

Two and a T

to resent s.-ith reacter pressure at 20 ps4 d4' been received in the control roon.Two or th ee days prior to che first depressu discovered in the conduit for the relief valve that-had inadvertentlyTne g

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opened.

i screw on the ccaduit ecver had pierced the insulation at

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cennection between the re ote cabling and the solenoid viring.

Q the nection was reinsulated and a smil The con '

burr on the end of the screw renoved.

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Between the first and seccad depressurization, all Target Rock relief valve solenoid pilot operators had been rebuilt.

r-af ter the second depressurization, the solenoid operator of the reliefUpon eatering the d valve that lif ted was found to be stuck in an internediate position so g

it was both bloving air into the valve air operator and venting air fro the operaer. A secoJ-ground was found to be caused by water in the n

Tne grow vas repaired and the air lines blova dry. solenoid housi All re~lef valve solenoid operators were renoved, and during bench testi it was found the solenoids were initially energizing at 90 volt ng holding cu rent'was reached. valve did ot drop out until a value of 6 to 8 volts DC and s, but the f.'

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take solen:1d operation susceptible to spurious groundsnese lov values of hysteresis c or phanton circuits.

, leakage paths, s

s Five of the eleven solenoids were found to have their 0-ring partially out of s

their seat.

could be recaptured in its seat by continued operation (approxina

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T tve:ty times).

in the ic11 cpen position.A burr on the plungers contributed to the 0-ring sticking g.,

Inspecticn of the valve solenoids revealed all to have dirt A

in the body area, and the lubricant us'ed to assenble the valves had turned black fron wp valve hez:.

the teflen _ ape had deterierated.Aust was found in several cf the valves, and at sone joints.

M Dirt was found around the solenoid f..

pilot in vari:us stages of being disledged f. cn their seatsseats and plungers, and sev

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The solenoid valves were cleaned and reasse= bled with new body internals and lubricant. A leak check of the valves indicated zero piston and poppet seat leakage, but all ten valves had some pilot seat leakage.

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j Af ter new plungers were installed, pilot leakage cas detected to be co=ing j

around the pilot seat and through porosity in the valve body.

Six valves v6.re made leak right by resetting the pilot seat, and four valves were rejected because of body porosity.

The latter were replaced with new g

valves.

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I The eleven solenoid valves that passed inspection were installed and functionally tested satisf actorily at 250 psi reactor pressure, and again at 930 psi and 20% power.

l The insulation of the Target Rock relief valves was nodified to naintain the air actuator and solenoid valves at a lover tenperature.

After unit startup, the solenoids were operating at less than 210*F, a tenperature saf e for prolonged operation.3 i

EXCESSIVE REACTOR COOLANT SYSTEM COOLDOWN RATE During the course of a routine shutdown for naintenance of the Oconee (t]

Nuclear S:ation Unit 3, when reactor power had decreased to approximately q.

15%, a systen transient occurred that resulted in the opening of a pres-R surizer relief valve.

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ir The power actuated relief valve had correctly opened when reactor coolant l

systen pressure reached 2255 psi, but failed to close when pressure dropped below 2220 psi. The open/close lights in the control roon did not indicate 5

that the valve was open.

As reactor coolant systen pressure dropped, the.

reactor t-ipped on low pressure, and the High Pressure Coolant Injection (EPCI) system actuated. Reactor coolant systen tenperature and pressure were 480*? and 720 psi, respectively, when depressurization terninated.

The initial drop of tenperature exceeded the allowable cooldown rate of LOO *F/h. by 1*F/hr.

The relief valve was stuck open because of heat expansion and bor'iS acid crystal buildup on the valve lever. The crystals dbed against the.

solenoid brackets and bent the solenoid spring bracket.

The valve was repaired and reinstalled.

The cause for malfunction of the valve position indication was not observed when the repaired valve was reinstalled.

Possibly, this alfunction could have been caused by the solenoid plunger

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sticking at s.ightly less than the full open position, or by crud buildup around the plunger-operated ninature control switch to the open/close L

lights.

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L Tne transient and associated events also caused the quench tank rupture T

disc to blow, mirror insulation to be separated from the bottom nonsle of

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the pressuriner, and the release of approximately 1500 gallons of reactor b

coolant to the reacter. building sunp.

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The release of coolant did not cause any significant increase of radiation h }

1evel in,the reactor building, and no radioactivity was released into the environment.

The excessive cooldown rate associated with the trannient was evaluated, and it was determined that the operability of the reactor j k and the health and safety of the public were not affected.

j k No other systen linits were exceeded.4

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% h; LOW FLOW FEEDWATER LINE SEVEPS AT 6x4 REDUCER 7 p 't N

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'inile the power level was increasing at Unit 2 of the Quad-Cities Station

.1 y ' *j af ter an outage, aid with both cain and low flow regulating valves partially 4

cpen, a feedwater vibration alars was received in the control roon.

The

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l unit was nanually scr-ed, feedwater pu=ps were tripped and the feedwater

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regulating station was isolated.

Reactor vessel icvel was controlled with

-,4 the Reactor Core Isolation Cooling Systen (RCIC).

Tne low flow feedvater line had severed at a 6-to 4-inch reducer on the downstrean side of the lov flow regulating valve.

Inspections also revealed 3E f0[

cracks in the low flor piping at the low flow riser junction to the main y 4 feedwater line and in the reducer upstreas of the regulating valve.

A The fricial cause of cracking was operational vibrations at the feedwater j

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- ir regulating station, and the break was attributed to vibrations at the j [

feedwater regulating station during transfer of flow from the low flow valve to the main feedwat.er regulating valve.

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.9 At no time was safe operation of the reactor threatened; all reactor 9 '5..

paraneters responded satisfactorily.

The total anount of water released h

S as a result of this occurrence was esticated at 12,500 gallons:

8500 d'

gallons fro = the severed line and 4,000 gallons from the service water

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deluge systen.

This water was discharged fron the site on a batch control basis, and activity at the release point in the discharge bay was less

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g than the Technical Specification limit.

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l-The low flow feedvater line had failed previously en June 10, 1974, when i

the low flow regulating valve ruptured.

This rupture, also, was believed Fl to have been partly caused by vibration during nor=al service, but the rain

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cause was i= proper cachining of the valve body for weld preparation.

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Corrective actions to prevent future recurrence include the installati a

of a' " drag valve" to replace ~ one of the main feedvater regulating valves on to provide more adequate flow control over a vider range of fl and reduce flow induced vibrations at the regulating station. ow conditions low flow control valve line is pinmed to be repiped to a i

Also, the path as another measure to reduce flow induced vibrations.gess rigorous

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UNFLANNID RELEASE FROM SITE BOUNDARY E

With Unit No.1 of the Calvert Cliffs Nuclear Power Plant at ste d l

alarn from the vaste area ventilation radiogas monitor. conditions a y state radiation monitor was also reading above normal.

The nain vent Investigation revealed latica syste=. gaseous radioactivity was being released to the auxiliary building venti The radioactive gas was leating fro:

volume control sa=ple bood..

a vaste gas conpressor and from the h following sampling, alleving leakage through a section of excessively The valve had not been completely closed the vaste crea ventilation systen. perforated surgical rubber tubing into the

,I vented the volure centrcl tank vapor space to the vaste gas systemThis cause Approxi=ately 46 Ci of Ie-133 and 5 Ci of Xe-135 were released duri incid ent.

release rate 14-it for noble gases.This release is less~ than 1% of the Technical Sp ng the.

inated while investigating the source of gaseous activityTwo individuals were slig readily and co=pletely deconta=inated.

, but they were not constitute an undue hazard to the health and safety of plant persennelIt w or to the general public.

The diaphra= of the vaste gas compr'ssor was replaced.

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been repeatedly perferatedsurgical rubber tubing on the volu=e control tank sa=ple poin The section of

, which had i=portance of regularly replacing used gas sanpliLy the gas sampling syringe, wa The end of proper operation of sample system s ng menbranes and tubing, radiation safety and chenistry technicians 1.ves was e=phasized to all plant RELEASE IN EXCESS ' F TECESICAL SPECIFICATION LIMITS 0

Over a. period of several weeks, containment structure internal p Unit 1, and it was decided to dr'liberately vent excess contain ressure I

n to the at=osphere.

pressure t

b

.1b !

,/

I

~

Based on radioactivity ceasure mts of the containment atmosphere "M

r release rate of 49,550 cfm would ensure co=pliance with Technical

,a Specifications.

would have resulted in the allevable release rateUsing the conta4*ent purg being exceeded, so it was decided t6 vent through the contain:ent purge -isolation valves without operating the fans.

tai nent pressure decreased to 0.05 psig.These valves were opened for four m Review of pressures recorded during venting indicated the actual release rate to be 51,300 cfm during the first ninute of venting, exceeding the limit by 4::

It was estinated the release resulted in less than 5x10' individual at the site boundary.

crem to an an undue basard to the general public.Therefore,, this. incident did not constitute During future centaincent ventugs, either one of the purge isolation -

t

+

valves will be throttled, or an alternate neans for more slowly venting the containment vill be provided.7 i

T?.ASSIER CF 2.: 0:. LING WATER TO CONTAINMENT 3ASEMENT During performance of a periocie test for safeguard syste= valve operatio at Unit No.1 of. the R.E. Gi mn Nuclear Power Plant, f

n established, and containment integrity was violated. refueling wa a. flow path from the An operator, while folloving a checksheet, closed valve MOV-8513 and erreaeously reopened it before the next step to stroke MOV-850B and initialed the procedural step "close MOV-8513".

for his hourly readings and recuested another operator to take them forHe then no him.

Returning to the procedure he saw that the next step af ter the last step he had initialed was to ope: MOV-8503.

with MOV-8313 open, flow was established from the RWST to sump 3.Upon the Upon receipt of alarts, the operator t=aediately secured the flow path.

It was estimated centainnent had been violated for approximately 3 minutes

[

There were indications rhat approvinntely 1-inch of water

!j containment floor.

the water was precessed according to normal proceduresNo danage to th e

g There was no danger to the plan The control room operater was reprimanded and, because of the naturer to th

,9 il this occurrence, prehautiens have been implemented so that control e of M

duties and routine cperations.Dpersonnel vill not have si=ultanecus responsibilit h

room j

m

j

<C l

9

.ta l

UE?2QUIF2D ACTU/JION OF D2RGENCY SYSTEY3...

)

With Unit No. 2 of the F.111 stone Nuclear Power St ti

phase, System (ESAS) was deenergized for maintenancethe cabinet for c "C" of the Engineered S fon in the power ascension a

a eguards Activation All other safeguards f

channels were energized.

noticed the positive logic power supply fuse light for ESASThe technicia vas out; this condition was indicative of a blown fuse channel "D" light bulb was replaced, but the bulb did not energiz The fuse indicator vas actually not bleva. removed, resulting in a loss of chenel "D" ESAS power becaus The fuse was then e.

e the fuse cf-4 logic condition was established, resulting in generatiWith chmnels "C" and "D" deenergized, a' 2-out-engineered safeguards actuation signals.

on of all shedding to occur from the e=argency buses. normal power signal art, with load Tae actuation of the ESAS co=ponents did not adver the plant or the health and safety of the publi sely affect the rest of c.

As a result of the ESAS transient, it was discovered an e=ergency bus was undersired.a po er supply for automatic closure of one e

generators onto period of about 12 ednutes.

Power was unavailable to this bus for a The "3" service water pu=p failed to start error, the pu=p was aligned to Unit 1.

Because of an adninistrative discovered in the water pump control circuit that prA wiring error was subsequently of the service water pt=p.

evented proper sequencing Also, two of the centain=ent air re irculation fans did speed.

The proble: vas traced to a loose relay to the control of b th not start on slow fans.

o All of the discovetec cagfunctions vere repaired and test d their correct operation.

e to verify 0

E e

E

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.; Vi-e

[4 REFERENCES h

Letter, B.B. Stephenson (Co=:nonwealth Edison) to J.G. Keppler, USNRC,

%{

1.

Office of Inspection and Enforcenent - Region III, July 3,1975. AOR J

No. 75-31, Docket No. 50-249.

• 3

{

2.

Letter, S. Surstein (Wisconsin Electric Power Company) to B.C. Rusche, USNRC, Office of Nuclear Reactor Regulation, Septe=ber 26, 1975.

Docket

)

No. 50-301.

3.

Letters, E.E. Utley (Carolina Power & Light Company) to N.C. Moseley, USNRC, Office of Inspection and Enforceme'nt - Region II, May 16 and p

?

21, 1975.

AOR Nos. 75-13 to 16, Docket No. 50-324.

E

?-

4.

Letters, W.O. Parker, Jr. (Duke Power Company) to N.C. Moseley, USNRC, W

Office of Inspec: ion and Enforcement - Region II.

June 27 and August 8,

{

1975.

AOR No. 75-7, Docket No. 50-287.

k 5.

Letter, N.J. Ralivianakis (Co==enwealth Edison) to USNRC, Director of I

Office of Nuclear Reactor Regulatien, August 27, 1975.

AOR No. 75-31, 7

Docket No. 50-274.

6.

Letter, A.E. Lundvall, Jr. (Baltimore Gas and Electric Co=pany) to J.P. O'Reilly, USNRC, Office of Inspection.'nd Enforcement - Region I, f

July 28, 1975.

AOR No. 75-44, Docket No. 50-317.

7.

Letter, A.E. Lundvall, Jr. (Baltimore Gas and Electric Co=pany) to J.P. O'Reilly, USNRC, Office of Inspection and Enforcement - Region I, August 20, 1975. AOR No. 75-48, Docket No. 50-317.

r-i E r

8.

Letter, L.D. White. Jr. (Rochester Gas and Electric Corporation) to

?

J.P. O'Reilly, USNRC, Office of Inspection and Enforcement - Region I, August 22, 1975. AOR No. 75-13, Docket No. 50-244, t

Y 9.

Letter, V.G. Counsil (Northeast Nuclear Energy Co=pany) to J.P. O'Reilly,

]

CSNRC, Office of Inspection and Enforcement - Region I, Septe=ber 18, 1975. AOR No. 75-17, Docket No. 50 336.

U i

I.

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Selected Safety-Related Occurrences Reported n.

in September and' October 1975 s=

CompiM bf V.'iiiimm R. Casto

.t:1 K

Sane of the oe..:::en;es :e;ar:ed in Jtiy and Aupst

~

i.i..

1975 seemed to be c:csegen:if eno;gh to revir.v: the reactor was opera:Ing at its low limit of 15'i of fa!!

thetsfore this section ns.c-i:ted frorn the ;:ert ;s power while being controUed completel.v automatie !Y i.a issue ofSuclc:rS: fen. Of tie c.:ca:en:e5 :e;; : edin and could not continue (clicwing the t*urthc:decress-E.!

September and O:te':er. Stee are :r.iewed he:e ing lead demand. At tMs point the opeia:ar ; laced the be: use of their ;ener2.,*nte:est tone !ert cpe:a:!r.s:

turbine control in mr.ua!. thus plaefag the control 5,$

i!

iD the trz.sien:s caud b:. the 0:once 3 eonta!

sys:em in the " load t:2. king" mode. This led to an rapid ineresse in the unit los:i d mand to s:.s:e:ns: C) the : ele:se of nc~:le gues 2:Zicn:2nd(3) automati:

5 zeedw2te I!ne 5:esk :: Quad Cities 2.

match the rea: tor power outpu:.1. the meantime the

~

7 mein steam bypass valves opened because of excess[i[

';i:.

rea: tor power; and, as CONTROL SYSTEM. CAUSES TF!A','SIENTS the main steam pressure G

dureau the valm eload.h conWI synem for me At Oconee 3. 2 feedw. iter flow to the steam generator could not follow twned : ' cre:a:e

• preisurized-w2:er rese:ct sF'.Viu by Duke Pouer Compmy r:e :

the rapid change in t.ni: load demand, and feedw$ter

!E-is; Clernson: S. C a t:r.iient o::urred whi:e the reae:n-Gow legged. This caused the feedwater Gow and the rewe: leve.

n was :q.ng de::esud fo: a ro:.ine r Inte.

steam. generator wate: !c.:: to oscillate, which in tura

- r nan e shutdown.8 - Prb: to the t=:::entiona:caused temperature and pressure transients in the i=..

s'ent. the res::c: ;ower was teirg: educed (:c= IOM

n.

,iE rea: tor coohnt system. V.~r.en the rea:ter-coclant-ih to 15 7 ofici!..wer by the cent:e! syne::.Whr:

sys:em pressure readed O55 psi, a power operated 7

vi fait p e.se:

15'

. 2 re.:;ted. the u:.i: Nd f.. 3.,,.

relief uire opened, as rq i:ed; bot th: ulve fai!ed to e5 SIV.tes. S:: the ;wer b.e.g gr ::.c..!

a.

e:o,e 2: ::(,o ;..

3, g.... aa, e.,,,, g,.,..e I!.!:

=

if 31Wien. T:, d: spa ity be: ween 2:

a oper. ;;. el @ts in the ee:.::,.,i room did not ini!.:re 2.i: 1:a d d:... d and powe: pne:2:.!c.. by the ::

E tha:

4.to: ex:n - te-c>e the vahe was still open. The reset. r oeiant a.,ct. A = 5

• EW. ve. 17.

• s 2. -a -r,-Se-a. im pres >; e continued down because of this open valve:

~

e g

~

,/

-: ~.

CPERATtMS EXPERIEftCES 107 C

E the rea' tor tripped on low pressure. and the high-NOBLE-GAS RELEASE pressure safetyinjection sistem autoc:atica2y actu-ated.

A calculated total of 63.7 Ci of radioactive gas was Although the operator elosed the lsa!ation va3ve on re'essed at Zion power station during venting of a the line with the fa!!ed postropers:ed re:ief va!ve mixed-bed demineralizer.8 This ststion has two PWRs

[

cwned and operated by Commonwealth immediately s'fter the reactor trip to ter:nhate the that are depressurization, the valve was reopened becat:se the Edison Company, Chicago,111. The maximum release wa:er level in the pressurher was risirs rapidly. The rate was calce:ated tobe 105,600 pCi/sec. and the rate iso ation valve was re:!csed when the reactcr coohnt w's es imated to have exceeded the technical-pressure reached 800 psi. A cooldewn cf 30l*F specifications limit of 60.000 pCi'sec for 6.5 min.The occurred during the first hour when 6e te=perature procedure for venting these demineralizers requires the I

t was below 5 0'F. The transient and associated events use of primary makeup water that contains no radio-j l aho caused the quench tank ruptu e disk, wid:h a:tive gases. However, this time the senting was received the blowdown frcm the powe -operated relief mistakenly done with the domineralizer connee:ed to i ~

valve, to blow. This caused the bruladon to sepa::te the rea: tor coolant system. This resulted in a direct

{.

from the bottom nozzle of the press:rizer, releasing puhway for releasing radioactive gas from the reactor 4

1500 gal of reactor cool:nt to the reacto:bt:ilding coolant system to the auxiliary building through

.l..

7 loose manhole cover on the equipment drain tank in sump.

'there was no simifica:t inc-esse h the :adiation the auxiliary building. Although the releese had no lesel in the reactor buildint, nor was a:y racioacthir's.

meas:rable censequence off site, the operating proce-

~

reb 5ed to the emircame$t. Also, the cooldown rate dures ic,r venting the dcmineralizers hase been did not affect the safety of the reactor.

strengthened, and the manhole cover has been tight-ened.

The power opera:ed re'ief va!ve st:sc in the o;en D calculated rele:se, based on the long-lived position because of heat expansiort, b:ildup of beric

,i ra i act:ve gases in the resetor coolant, was about 1 Ci l

acid crystals on the valve levei, rubbhg of the lever f mixed notte gases. However, the noble gases with against the solenoid bra-kets, and lending of the ve:) short hdflives in the coolant system were also solenoid spring bracket.

released. An attempt wi!! be made to determine the The valve ias repa:. d and rem.nalled. 2.d the

.,r re quzatity of these gases more accurately.

!t

se pesttion. di:stor c.ested up.

prcblem w.th the va.

in i

The fo!!owing corrective ac-ions have been ecm-pleted:

FEEDWATER LINE BREAKS

1. The unit shutdown procedures for a Oconee ON BWR units have been revised to include a dunge that will Xt the Quad Cities Nt.c! car Power Station, cwned l!

prevent decre:s:ng unit loaddemand be.ow 10.\\1W(e) by Commonwealth Edison Companv. Chicago 111..

before placing the control systemin the trackSg mcde.

tait 2 suffered a break in the body of a 4. by 6 in. '

~

r This minim!zes the error be: ween the teit load dernind reducer on :he downstream side of the low flow and generated power and redu:es the postatety of

,hr vdve in the feedwater system.' Unit 2 was feedwater flow and reacter coo' ant syste:n ::::si nts.

e cmine u

~

2.The power a:tuated - pressure relief valves of 170.\\1h(p in power after an outsee and was producing e) at the time of the incident. Operators were U:its I and 2 will be inspe:ted as soa: as p:ssfale for on the scene observing the transfer of Gow from the

[

2ny indication of buildup of bori: a:id :rysta:s.

joy,ge,. feedwater reculatine valve to the main feed.

3. To verify. the proper functicr.ing' cf power-water regula:ing valve because problems in vib:2: ion actuated pressurize relief valves. they will be yded urine this cperation had been execdenced. A feed-prior to startup with a test s!_ mal corresp:nding to nter sib:ation alann sounded when both valves were 2'35 psi.

partially opened. The unit was manually scrammed 4.Tne quench tank rupture disk ns rep *ac=d, the when the low-flow line just downstre:m of the bcttom nozzles on the p:essurizer we:e dye peset: ant low. flow regulating valve started to seser.Not unh did tested.and the insulation was repia:ed.

the line break, but crseks occurred in the lew hw

' S. Operating personnel were advised of titis in:i-piping at the low Cow riser junction with the main dent and given specific instru:tions to ir=nediately feedwater line and in the reducer upstream ci the

,,]

close the isolation vahe.

regu! sting valve.

j

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PAGE 23 PusCI:SS LD luvul uG JUL1 TAC 1 LIT 1/

SYSTLn/

EVEkT b!TE/

(

COntONENT/

DOCK ET NO./ kEPONT DA2E/

CAUSE CODE (UNT!iOL NO.

REPOhT TIPE M ENT DESCitjPTION/

CAUSETbESCu 2 PTl hm ICONhE-3 050-02fs?

061375 110-75-7) k1CESSIVE IlEACTON COOLANT SYSTt:n rootDOWN laave.. huh 3N4 1 CTllEk COOLANT Sur:SYS 4 CONTHOL ' 01207t.

062775 NE SiluTuuMN (IM IOWFH) A l DV Lil.1CTUATED l!!tSSUhl% Ell llE!.3EF VALVY Or V A LVE OPEh&TultS 2-VEEK Tile N llE?lh1NED Ol'EN UNTAL AN ISOLAT10N VALTE WAS Shut. t ilt 101 LEG l'ERSONNE!. EkilOk CoutbOW N JN Tilt FIRST 130uu Eact:EbLD Till: 100 DEGllLI: r LIn17.

PEhSONNEL th holl: Puon t-Ttli OPEft&TOk A CT *.O N COULis II AV E Ph LV ENTEls

t. ACES $11tl COO L1,0 v ll. IIEL1).P V ALVE W AL llEl A318ED (hu!!!C ACID DUILUUl* ON O! EllATQal.

CCOWLE-3 O Su-02 tl7 061975 IAO-72-u) REACTOk uultp1NC LNt:' S t.EHED S AFECU A hDS PkCSSuht TkAbsr.1TTER PO C ;TNis NT ISOL ATION SYS

• CONT 012UU ts 07037$

UND OUT OY CAL)hkATION (UY 6.b PSjC) buhlNG h007]NE ILSTlWG LHhshG CbLb IN STR Uf5 t:NT AT] ON 4 CO b7H O LS 2-WEEK Silu TDO W N. Til t 2 I., 3 ACTu ATION LOGIC WAS NOT ArrrCTt:D.

Co r:PO.41.WT FA3LUME COnPONENT FAILUIIE: SETPOINT Dil1 FT.

O YSTFil Chr.tK-1 050-0218s u h u'F L

( AO 'J b-I t.)

SulVEJI.LANCn TI.SrtNC ON Tilt

• S ELECTuon1 TIC alELit.F V A LV E PH ES t e E14 G CuHM COOL]NG SYS

• CON 1 b12h22 06247S ult e SWITCill S Willtt: lilE htACTuh VAS 3 N Ti! E Il Fru t t n0!s P uhV tall b 2 SVlW

• l h STH Dill:NTATION 4 CONTh0LS 2-V t.l:K S ( I A O.lb 1; 1 A:! JI') TO Tit 1 P Ar 5 A:du 3 PSIC AbOVE TllEllt SAI! nun ALLOWAhLE COnlONENT PAllukt '

Tid A P 3*01 ETS or 1009 AND 100 *e PSIG.

I N ST h un t'NT SMT IO)MT blt ] F T.

SWITCllES W Elf E Ill*St:T TO ALLOWahlh I.t V E L*.*.

OYSTEh CitEEK-1 650-0219 o f. WIS 110-75-17) leuu1NG S itit V EI LI. A NC E, Cult h LPN A Y SYSTEM PAHALLEL JSOI.ATION VAL EcEltC Cultt: COOLJNd SYS

• CONT 0121st!9 062775 VL ( V-2 0 - 15) P131.1D TO CLOS E.

A DhuntN MOTON chrAnt.k STAh WAS DIPLACID A C11tLUIT CLOSENSf1 NTr;HitularElls 2-WEEK N D Tilh V Al.V t; T MNT a ll N A T 4 S P A Uf tsH a l.Y. A NPuuNDANT VALVt. WAS t,PtHAltLE.

Con Pu N EN T P A l l.U h at Tlt r 'l' A li tale Tilt "h*

l II A S L D I" Tit P V7LWF f0 TON hhhAnth ST Ali W AS Dhtsd tN NP.SU LTING IN IN I'Eh f11TT!;h r IsUS lif.It CO N T A CT. '.*ll e LOSS OF UNL l il A S I: C A US t.b l if t OYSTER citer:n-1 V ALV L abruk TO tis l P.

050-0219 Ob2f1$

(Au-75-1-) buu llu; OPER A1 t uN, TWO il A NDilOLE COVEHS IN Tilf. STANbHV GAL TkEA CNT3n3T Cut. bus t:ES CONT 3tOI. SYS D 120'J is 670175 TMErlT FILTEh TilAIN W rlt h stOr IN I LACr.. The COV EuS Wr.h u InnLbl ATI.L1 FILThh5

• / -W t FK b IN ILACh. Tilj S is L leu C t:b Tl! M A bl L]TY UP 'Illl 5 SIlGTS TO FUNC110N PllOPl li L I.

SI'.uun LEPLCTIVE Ph0CUtrohtS A h t:b u k ts A N T Tll A I N WAS AV A X LA tlLI.

u se U L'l Ele n 1 N E b. PhDCE bu p e.S FalH FI LThlt TISTJNG WP:ItL ltEV ASED.

P A LIS A bl S-1 0S0-0255 060375 A Htvitw OF CI:1.Oll' NATION OTutu Aur W A rt:n SrS. Cha TLu tS D irn.s oo u?S Sr:.Tre. suoWtu tuAT r. S...., :. rn64ucG.i.4.u.i.iu Nul un S I:n t W rLv nl.1 A lh PA T.1 t.NT bP Tile CIDSE D C f CLE Cu r t.t.Nat t. CubLI NG UThl.fi CbtlPub MNTS f -W I;E K his SucL s.l;OU1h Lb f1E ASUh tP t.NTS N f:V tX TAKEN. CL68. l N U lil!.dll a ht;l' lA10 LAAL 8 DEYt.CTI V ). l'huCEDu tt ES

] Cll } G A N C6 hlb ilAVt: !!EACHED b.0b PPn CsJnP&ltED TO T.S. Lilll? OF 6.be PI n.

( A 4 -7 *.- 1.').

laur ACCOUNT]hb YOh COk ris!!surinh Fh0A COOLING TOWth. PleOCl*DLtle r Pl.lslit ta k,

A N L, 1s LC II;;41 NG OPEl57]uW OP.hEWLY INST ALLED Clu* l.b CICLE CuteLa mi. SYSTI.fl.

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Reference 7 y%yty

- '=4_.

,e M C.* UNITED STATES

• * - I-l -

nub. EAR REGULATOR. Y COh..-...

~~

S,,

REGION !!,',

230 PEACHTR EE STR CCT. N. W. SUITE S 15 In Reply Refer To:

. A0G 2 71W5

.f ATs.AnrA. C CO RCI A 30303 2

IE:II:TNE 50-269/75-9 50-270/75-10 50-287/75-10 DUPLICATE DOCUMENT

.!!r Entire document previously k-entered into system under:

Duke Pcuer Cc=pany ANO QCof0W0%

7 ATIN: Mr. h*1p? 6 O. Parker, Jr.

43..

Vice Presistent.

N o. of pagea:

, i~.

Stenn Production P. O. Box 2178 422 South Church Stn eet Charlotte, Korth CarM N 28242 Gentlemen:

This refers to the inspection conducted by Mr. T. N. Epps of this office on July 29-31, 1975, of activities authorized by NRC Operating License Nos.

[

DPR-38, DPR-47 a:d D7R-55 for the Oconee 1, 2 and 3 facilities, and to the discussion of our findings held,vith Mr. J. E. Smith at the conclusion f

of the inspection.

,i.

hi Areas examined dcring the inspection and our findings are discussed in the y

enciesed inspection report. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews j.[

vith personnel, and ebservations by the inspector.

p.;

'li-We have examined ar tions you have taken with regard to previously reported

![.

i-[k unresolved ite=s.

Ihese are identilied in Section IV of the sucrary of the enclosed report.

!?-

During the inspection, it was found that certain activities under your JE license appear to be in concompliance with KRC require =ents.

These items

[

and references to pertinent require =ents are listed in Section I of the i

sn-mry of the enclosed report. Additional inforr.ation on Abnorral Occur-rence AO-237/75-7 in your letter dated August 8,1975, was received by

!E this office. No additional response is requested.

IE

21:

~

In accordance with Section 2.790 of the NRC's " Rules of Practice,"

=

Part 2, Title 10, Code of Federal Regulations, a copy of this letter 4E and the enclosed inspection re' port will 'ce placed in the NRC's Public Document.Roon.

If this report contains any information that you believe

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to be proprietary, it is necessary that you cubmit a written application to this office reques. ting that such information be withheld from public r

disclosure. If no proprietary infort:ation is identified, a written statement to that effect should be submitted. If an application is submitted, it I:ust fully identify the bases for which information is e1=4-=d to be proprietary.

The application should be prepared so that infor=ation sought to be wit *nheld is incorporated in a separate paper and referenced in the application since the application will be placed in the Pcblic Do u:nent Room. Your application, or written statement,.

should bs subgitted to us within 20 days.

If we are not contacted as specified, th6 enclosed report and this letter may then be placed in the Public Doctent Roon.

~

Should you have any questions concerning this letter, we vill be glad to discuss thc= with you.

I Very truly yours, i

i

/

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Norman C. Moseley Director Enclosure.

IE Inspection Report Nos.

50-269/75-9, 50-270/75-10, and 50-287/75-10 i:

cc:

J. E. S=ith, Plant Manager Ocence Nuclear Plant i.

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NUCLEAR REGULATORY COMMISSION

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REGION 18

.j : ' ~

22o PtAcHTacu sinscr. n. w. surre sis AT L.ANTA. C E O R C I A 30303 yh IE Inspectica Report Nos. 50-269/75-9, 50-270/75-10 and 50-287/75-10 Licensee:

Duke Power Company Power Building 422 South Church Street Charlotte, North Carolina 28201 J'

~

Facility Na:e: Oconec Units 1, 2 and 3 Docket Yos.:

50-269, 50-270 and 50-287 License Nos.:

DPR-38, 47 and 55 Category:

C, C and B2 Locatio::

Seneca, South Carolina Type of License:

B&W, PWR,' 2568, Hw(t)

Type of Ins ection: Routine, Unannounced Dates of Inspection: July 29-31,1975 r

Dates of Previous Inspection:

l' y 27-30 and June 3-6, 1975 Principal Ir.:pector:

/

pf c. -

8-A7 '7.I

' Inspector Date j

T. N. Epps, Rea e Facilities Oper~t ons Branch Accc=panying Inspectors: None *

\\,

F. J. Long, Chief C.T _

O._.

3-Revie. red by:

sm

/ D&te Facilities Operations Branch

• See Details II I

hp DUPLICATE DOCUMENT hd.

Entire document previously r

entered into system under:

C###

ANO $@l6,8[/)9h

(

No. of pages:

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IE Rpt. Nos. '50-269/v9, 50-270/75-10.

'g.f ;;;

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!E

. and 50-287/75-10

..; :~0-- v.

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SUMMARY

OF FINDINGS

~

~

T

{

NY I.

Enforcement Ite=s A.

Deficiencies

1.. Contrary to Iechnical Specification 6.6.2.1.a, abnormal occurrence report A0-287/75-7 did not include an analysis and evaluaticu of the safety implications involved in the blowdown of_ the Unit-3 reactor coolant systen nor did the report address the causes and corrective actions caken to prevent recurrence of the incident.

(Details I, Paragraph 2)

(Unit 3)

/

D II.

Mce:see Action en Previously Identified Enforcenent Matters Not inspected.

III.

New En-esolved Ite=s None IV.

S ta tts of Pravio: sly Reoerted Unresolved Ite=s Oconee 1, 2 and 3 - (50-269, 50-270 and 50-287) e 74 -10, 03, 11/7 NSRC Review Caoability T*nis iten is c1csed.

(Details I, Paragraph 6) 75-3/1 Analvsis of !.icuid Wsce Samples Ihis ite: is closed.

(Details II, Paragraph 2) 74-7/2 Activity _in :he Cocoenent Cooling Systen This iten is closed.

(Details I, Paragraph 7)

V.

Othe-Significant Fi dings None

~

E VI.

Manarece,nt Interview A m agenent interviev was held nn July 31, 1975, with Mr. J. E. S_ith and ne=bers of his staff. Ite=s discussed included the nonconpliance iten in Section I of the su=:ary of this report, surveillance testin3, m

two enresolved ite=s in this su==ary and settlenent of Class I

1-stru:tures.

U Further discussiens vere held with licensee corporate canagement on August 5,1975, concerning additional infor=ation on the Unit 3 blow-down that occurred on June 13, 1975.

5..:.

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IE Rpt.'Nos. 50-269/75 '9 [ 50-270/,75-10 '.

._l_ ~~ f--

~~

and 50-287/75-10 x..

y-

. =..

DETAILS I Frepared by:

/M "'

8L/'78 T.N.Epps, React /rInspector Date p.

Facilities Opera,ti ns Branch

=

Dates of Inspectiont July,29-31, 19 75 I:

.8 J. C "~ -

Reviewed by W

F. J. Long, Chief f Irate Facilities Operations Branch f

r 1.

Individu.als Centacted Duke Po er Cc sany (DPC)

.l J. E. S=ith - Manager, Oconee Nuclear Station

[~~

J. W. F*

ton - Director,.Mni' istrative Services L. E. Sch:id - Operating Superintendent O. S. Bradha:2 - Maintenance Superintendent R. M. Koehler - Technical Services Superintendent T. S. Barr - Tech:ical Services Engineer R. P. Bugert - Trair f eg Supervisor

..r i

2.

Unit 3 RC3 Blowdoen

,I I

Oconee Technical Specification 6.6.2.1.a requires that written abnormal occurrence reports describe, analyze and evaluate safety inplications

[

and outli:e the ccrrective actions and ceasuus taken or planned to prevent recurrence.

Contrary to the above the licensee's abnor:a1 occurrence report i

(AO-237/75-7) did not fully describe, analyze and evaluate safety inplications and cutline all corrective actions.

The licensee's report i

primarily addressed the excessive cooldown rate of 101 F in one hour j.

rather than addressing the entire reactor coolant system blowdown and l'

the safety implications of the incident.

i Apparently the initial transient was caused by a transfer of the turbine into canual while the unit load demand (ULD) was at 65 MWe y

and reactor power (auto =atically controlled) at 115 MWe. This eventt'n11y caused levels in the once through steam generators to swing, es causing ECS te=pe ature, pressure and power swings.

RCS pressure

.=

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.~

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=

IE Rpt.' Nos. 50-269/75-9, 5 0-270/75-10 I-2 and 50-287/75-10

=

spiked to 2267 psi which caused the power actuated relief valve, on the pressuriner, to open. The valve nalfunctioned and renained in the open positic=. In addition a solenoid operated plunger that actuates position indication lights in the control roon for the pressurizer relief valve r.alfunctioned.

As a result of this incident reactor coolant systen (RCS) pressure E

decreased fran 2250 psi to 720,21 s-ithin 26 minutes.

The reactor 1

tripped at 1800 psi a:d high pressere injection (HPI) initiated at-1500 psi.

f The transient was ter-f r ated then a block valve was closed isolating I

the opened relief valve.

During this transient the rupture disc in the quench tank ruptured due to stea= from the relief valve building up pressure in the quench ta:'.:.

App:o cirately 1500 gallons of prinary water were lost through the que=ch ca:k to the containment.

Insulation 'on the bottom of the pressuriner was da: aged t./nes the rupture disc ble-The licensee's report did not address the initial cause of the transient and cor active actica to prevent recurrence; why the block valve that isolates the press ~urizer relief valve was not closed sooner; corrective actica to prevent recurrence on all 3 Oconee. Units of the proble:s wit ~c the pressurizar relief valve and position indication equipnent; possible dacage to the pressurizer; or activity released.

The inspector stated to Ococae site person 2e1 and later to Duke Power Cu._,.any corporated perscanel that whenever rapid uncontrolled depressurization cf the pricary systen occurs causing HPI initiation and loss of soa pricary coolant, abnornal degradation of the pr'-=ry coolant b-studary has occurred even if blowdown is through an isolable fault if the fault is not isolated.

l The licensee agreed to subrd.t supplemental information on this subj ect.

3.

Surveilla:ee The inspector reviere several surveillance testing procedures 5

and rasults inclufing the follot.-ing subject areas.

f r

RCS Cha-htry E.

RCS T n'. a ge

[:{

Centrol' Rod M:vesant i?;_

ic

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%= m.=.65#5hi^& = **E WE I-3_. f

... ' '. 2.5Y"s!U 'f.t "

2 5

IE Rpt. Nos."50--269/1529.; 50:270/75;--l.Q

~~

"75 2

'.7 '

Ef.

and 50-287/75-10... 4. = - "- - -

1.

~

g.

. c..

e Emergency Feedwater Pu=p Testing

@i Secondary Coolant Activity 4: '

Spent Fuel Fool Water Samples k

Electrical Systems lE HPI and LPI Pumps

[.j,3 Some Reactor Building Local Leak Tests

!E C::.

?!

Within the scope of this review no noncompliance items were E

i identified.

E g

4.

Operator Refu.alification Program

..E

..:z 2'..]. l, A licensee representative stated that NRC licensing personnel reviewed some operator requalification e.xa=inations that were given at Oconee.

iii:

==

The requalification program received final approval June 18, 1975.

.E E".

5.

Settlement of Class I Structures 1:.

The inspector inquired as to whether the licensee has a progras

'Cl..

for maasuring differential settlecent of class I structures, such E

as, the reactor building and ' auxiliary building.

The licensee El stated that such a progran does not exist at Oconee since 7

the facility is built on solid rock.

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6.

MSRC Reviev Capability G:

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The licensee furnished information to the effect that one permanent E.

cenber of the NSRC has an M.S. degree in Materials Engineering and z.

provides capability for reviewing netalurgical considerations.

.h This ites is closed.

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7.

Activity in the Component Cooling Systen

=L The licensee's letter to the NRC's Region II office dated b

May 9,1975, stated'-that codification had been installed which i= 1 added additional isolation valves between the component cooling

):e )

drain ' cant pu=p discharge header and the niscellaneous waste transfer E..

pu=p discharge header.

The level of activity in the component cooling E-systes has decreased *; The licensee stated in the letter that i:: l monitoring of component cooling system activity will continue until E

the activity decays to nor=al background. This item is closed.

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Binghis P=cs 3olts Tne 1:spe= tor questioned licensee personnel about testing done on Binghzu Pc=p hold-dova bolts. A licensee representative stated that duri:g ISI baseline testing these bolts wera UT tested and found to *ce acceptable.

Samples of these bolts will be retested at regular intervals such that all will be tested in the 10 year ISI cycle.

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DETAI II Prepared [

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['2 W. L. Britz, Radiat'i'on Specialist Date

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Environmental Protection, Materials

. Radiolb31 cal Protection, and h

Special Proj ects Section Radiological and Environmental c

Protection Branch b.

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• Date of Inspection: July 22,1975 y;

Revieued by: $/., ba.lc m f S/.z 1/77

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R. L. Bangartf/ Senior Healtn Physicist 'Date

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Environ = ental Protection, Materials

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Special Proj ects Sdetion

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Radiological and Environmental E-Protection Branch E:

1.

Indiriduals_ Contacted E;

1 J. W. Hanpton, Director, Administrative Services (Acting Plant Manager)

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D. L. Davisen, Assistant Health Physics Supervisor 2.

Analysis of Liould Waste Sacoles (75 3/1)

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A.

Ihe licensee.is required to neasure quantities and concentrations of radicactive caterial'in effluents froc his f acility. During c

previous independent measureuent checks of June, Septe=ber, and October,1974, the licensee's ability to neasure radioactivity

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in test standards and plant effluent split samples was evaluated.

F Some results of the licensee's ceasurements of gamma enitters and strontita in liquid were in disagreenent.

It was also determined 5

that gross beta analyses had not been normalized against results f

of total isotopic analyses when used to determine values for reporting releases of liquid effluents.

See IE Report Nos.

50-259 /75-3, 50-270/75-3, and 50-287/75-3.

3.

On P. arch 18, 1975, liquid and gas split samples were collected by the Division of Fadiological Health, State of South Carolina and

,5 analy=ed by the, licensee's laboratory and the NRC's reference labora tory. There were eighteen ceasurenent comparisons. Twelve co=pa:-isons were in agreement, four vere in possible agreement, i.E and two in disagreecent. The disagreements were on antimony-124 E

in the liquid sample and krypton-85 in the gas sample which vere k

not detected by the licensee, but were reported present in concen-gp trations greater than 10% of 10 CE2 20 Appendix B, Table II, by g,

the NRC's reference laboratcry.

It appears that these two isotopes E;

vere not detected due to the short counting times,used.

The gg.

licensee has co=nitted to count future split samples for about E

one hour to achieve lower sensitivities. The gacma esitting neasurecents are now resolved.

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• The inspection action was an in-office evaluation of analytical results, which were discussed by telephone with the licensee' representative on July 22,1975..

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On April 3,1975, strontius test standards in a liquid sample and on a particulate filter were sent to the licensee's labora-tory "for m1ysis to resolve the disagreenents of June and -

S ep te=b er, 1974. The licensee's laboratory procedures were al.so reviewed by the NRC's reference laborawy.

Cocnents provided to the licensee's laboratory by the NRC's laboratory included:

use of Sr-85, Ba-133, and Y-88 as gr m tracers to check various steps in the procedure for removal or yield Ectors, more exact control of pH, filtering rather than c*ntrifuging in one step, and controlling temperature in anoth(r step.

The results of four measure =ent conparisons for Sr-69 and 90 were three agreements and one possible agreecent. The previous strontium measurements on the March 181.iquid analysis were also in agreement. The strontius

=easr..ecents are now resolved.

D.

The licensee in a letter of April 11, 1975, reported he has acv determined ad is using a normali::ation factor, on the gross beta calysis, based on total isotopic analysis, when naking radicacrive liquid releases based on the gross beta an.alysis.

This itec is no.i resolved.

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.,4 Mr. Hor =an C. Moseley, Director x.,4;p.;4,,, ;y,.;j U. S. Euclear Regulatory Cocaission

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230 Peachtree Street, Northwese Atlanta, Georgia 30303 ke: Oconee Unit 3 Docket No. 30--287

Dear Mr. Mose' cy:

Pursuant to Sections 6.2 and 6.6.2 of the Oconee Nuclear Station Technical Sp'ecifications, please find attached Abnormal Occurrence

{

Report Ao-287/75-7.

Very truly yours, s

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Willian 0. Parker, Jr MST:vr At tach =est ec: M. Angelo Giachusso DUPLICATE DOCUMENT Entire document previously entered into system under:

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OCOSEE UNIT 3'.

Entire document previously entered into system under:

Report No.':

.m-287/75-7 V Md.h[ M Of ANO Report Date:

J=ne 27, 39 o-No. y of pages:

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Occurre-ce Date: J:=e 13,1975

• Facility:

Ocente V:ic 3, Seneca, Scuch Caroi b Identificatien f Occcrrence:

Excessive Reactor Coolant System cooldown rate Conditions Prior to Occtrrence: Shutdows in pro'gress Descriptien o Cccurrenca:

On June 13, 1973, a routine shutdown for =aintenance was in progress on Oconee Unit 3.

• 4he: reactor power had decreased to approxinately 15 percent, a sinor sys:e= ::ansient occurred which resulted in the opening of power-actuated pressu-ize: re ici valve 3E::-66.

Valve 3RC-66 remained open and a Reacter C:olant Syste depressurizatica cc:tinued until isolation valve 3RC-4 was shur. The Reactor Coolant System te=perature and pressure were

^480 F and 720 pri, respectively, whea tha depressurization was terminated.

The shutdev: ras cc:tincad. vith a coaldovri rate of 100 F/hr as specified

.c in Technical Specification 3.1.2.3; however, when the initial drop in te=perature d=e to depressurizatica was co=bined with the subsequent cooldova, the c:oldern r' ate for the first ho:: vas 10l F.

O Desienatien of locaren: Ca :sa of Occurrence:

The apparent c2:se cf this cecurrence was operator error, in that the operator did not conside: the #,'-1.al RC te=parature drop, which occurred during de-pressurization, whe: astablishing the subseq:ent cooldova rate.

The reason RC-55 remained c;en was due to baric acid crystal buildup on the connectinji; pic cf the lever arm of the pilot valve.

In addition, a solenoid-cparatai plunger was stuck in the open position.

Analysis of Oce: rer:e:

This incident resulted is exceeding the al2.orable cooldown rate of 100 F/h.-

.by ICF/hr. D e to the

• design conserv/ tris: of the reactor vessel, and transients sich have prav'.*ously beca scaly:sd, it can be concluded' that.

the health a d safety of the public vas not affected.

Corrective '--*cn:

In the future after such a transient, an evaluation vill be perforned to deter =ine tha - M allevable cc.oldova rate to be utilized.

Valve 3RC-66 e

was re:nved, repaired, and replaced.

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August 8,1975

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Mr. Nor=an C. Eoseley, Director

0. S. Nuclear Eegulatory Co-hsion Suite 818 230 Peachtree Street, Northwest DUPLICATE DOCUMENT

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Atlanta, Georgia 30303 Entire document previously entered into system under:

l Re:

Oconee Nuclear Station Docket 1;o. 50-287 ANO bkM No. of pages:

Dear Er. E:

seley:

My letter of Jcce 27, 1975, transmitted to you Abnor=al Occurrence The following infor@mation provides additiotal infornation relating to Report AO-287/75-7, cessive Reactor Coolant Systen Cooldovn Rate this occurrence and associated corrective action.

As stated in A0-287/75-7, when reactor power le rel had decrease'd in the course of a routice caintenance shutdo.n, a nicor systen transient occurred, vhich resulted in the opening of the power-operated xelief l

valve 3RC-E6. Prior to the systen transient, reactor power was being l

reduced fren 100% ?? to 15I FP in an orderly canner by the Integratad

' Control Syste=.

7:en 15ll: FP vas reached, unit load de=auti was 65 ewe, l

and power generatica vas 115 h7e. This difference between unit load l

de=and and pover generation existed because the reactor was operating at its lov linic of 15: F? while in auto =atic 1CS' control and could not further follov unit load de=and.

Meanwhile the control operator placed the turbine control st2 tion in rmm =1, leaving the ICS in the " load l

tracking" code.

Tais led to a rapid increase in unit load de=and to i

catch the gene ated cegavatt output.

In the meantice, the main steam bypass valves opened; and when the = sin stea= pressure decreased,. the valves closed.

The ICS control of feedvater flow could not follov the rapid change in unit.lcad de-=nd and steam pressure; consequently, feedvater flow and stean generator level oscillat'ed, resulting in the Reactor Coolant Systen te=perature and pressure transient.

The power operated relief valve, 32C-65, o,:ened when RCS pressure reached 2255 psi but-failed to close when the pressure dropped below 2220 psi, although the open/ closed lights in the control room did not indicate that the valve was open. Consequently, RCS pressure dropped, the reactor tripped on lov press'ure, and the EPI system actuated.

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Altbough the operator closed the isolation valve, 3RC-4, i= mediately after the reactor trip to ter-innte the depressurization, the valve was reopened because of the rapidly rising pressuri=er level.

Valve 3RC-4 was finally closed when RC pressure reached 800 psi, terminatidg the pressere transient.

The subsequent controlled cooldown of the Reactor Coolant Systen, when ec=bined with the te=perature drop during the 0

transient, resulted in a cooldown of 1017 during the first hour when te=perature was belov 530 F, contrary to the provisions of Technical Specification 3.1.2.3.

The transient and associated events also caused the cuench tank rupture disc to blev open, Mirrqr insulation to be separated from the bottes nozzle of the pressurizer, and the release of approximately 1500 gallons of reactbr coolant to the Reactor Building su=p.

The release of reactor coolant did not cause any significant increase of ndia ion level in the Reactor Building, and no radioactivity was released into the environnent.

As addressed in A0-287/75-7, the excessive cooldown rate associated with the transient has been evaluated and it was determined that the health and safety of the public vas tot affected. No other system limits were exceeded.

The failure of 3RC-66 to clos'e and the calfunctioning' of the valve position ind. carica in the Control Room have been investigated.

It has been found that the valve was stuck in the open position because of heat expansion, boric acid crystal buildup on the valve lever',

' rubbing of the level against the solenoid brackets, and bending of the solenoid spring bracket.

The valve was repaired and reinstalled.

The calfunctioning of the valve position indication was not observed when the repaired valve was reinstalled.

This malfunctioning was apparently caused by the sticking of the solenoid plunger at slightly less than the full open position or by the crud buildup around the plunger operated nicroswitch to the open/ closed lights.

Additionally, to prevent recurrence of this incident, thefollowing corrective actions have been or will be implemented.

1.

The unit shutdown procedures for all Oconee u' nits have been revised to include a change that vill prevent decreasing unit load demand below 120 INe ~oefore placing the ICS in the tracking mode.

This would nininize the error between the unit load demand and generated.

po.er and thus would reduce the possibility of feedvater flow and RCS transients.

2.

The Units 1 and 2 power-acchated prescurizer relief valves vill be amined as soon as possible for any indication of boric acid crystal buildup.

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3.

To verify the proper functioning of RC-66, a test to cycle RC-66.

Prior to startup vich a test signal corresponding to 2285 psi will be incorporated into the station operating procedures.

4.

T"ce quench tank rupture disc has been replaced, and the bottom nozzles on the pressurizer were dye penetrant tested and the Mirror insulation replaced..

5.

Operating personnel have been advised o5 this incident with specific instreetions that ir=2ediate closure of 3RC-4 is the proper corrective action for such an occurrence.

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• truly yours,

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