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Crystal River Unit 3 was in PODE 5 (CDI.D SHUI'DOWN) fran February 27, 1989 toJune 1, 1989. Durirg this outage, NRC inspectors discove ed deficienciesrelated to envirw isital qualification of plant equipnant. S e p ntinvestigations of envite.ustA qualification records have discoveredadditional deficiencies. Deficiencies included improper cables and splices,inproper silicon oil level in instrument junction boxes, and problems relatedto valve motor operators. Problens were the result of deficiencies in detaileddevelopnent and inplementation of the envite..sital qualification program.,

'

Utility penminal have repaired identified envire..ard qualificationdeficiencies, or have justified continued operation with the deficiencies untilrepairs are ocupleted. 'Ihe utility has enbarked on a major voluntary effort toreview the existirq Envile. td Qualification propath, and to correctadditional envim isital qualification deficiencies that may be discovered.

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Crystal River Unit 3 was in MXE 5 (CDID SHlTfDOWN) frun February 27, 1989 to |June 1, 1989. Ikaring this outage, a Nuclear Regulatory Ocamission inspection |

l team disoevered several deficiencies related to environmental qualification| (BQ) of Plant hWnt. Inspectors fourd the follwing deficiencies: ;

'|

| 1) Igroper electrical cable (CBL) and splice (CON) installation, !

including cables and splices not qualified for subnersion found Ilocated belw the Reactor Bailding (RB)(NH) flood level, !

)

2) Igrger oil level in instrument junction boxes (JHX),

3) Missing or painted over T-drains (mW), and missing or capped greaserelief fittings on valve motor operators (84),

1

4) Deterioration of wires and grease ==wiated with the Pilot Operated jRelief Valve (PORV) (AB,RN) Block Valve (Tag No. RCV-ll) [AB,SHV). )

i

UNOUALIFTEn CAmm fCBL1 AND SPLTCES f 00N) || On April 26, 1989, during the NRC Envilu =uiuti Qualification (EQ) Program |

audit, inspections found that cable splices on signal cables frun two pressure I

transmitters (PT) had not been installed in accordance with the splicemanufacturer's application guide. 'Ihe application guide required that eachsplice band radius be no smaller than five times the outside splice dic. ster. '

'Ibe manufacturer had no data to determine whether or not splices could bequalified with smaller bend radii. Inspectors found bend radii that were onlytwo to three times the outside radii of the splices. 'Ihese splices wero

i

located in the Reactor Coolant System (AB) instrumentation Wiring, betweenconduit seal assenblies and the field cables.

1

Original plans called for installation of junction boxes between the ,

instrument conduit seal assenblies and the field conduits. 'Ihese boxes were tc, I

be large enough to allw splice installation with acceptable bend radii. Due J

to seismic mourting difficulties, the plans were revised to specify 3/4 inchcoMulets (CDT) instead of the junction boxes. 'Ibe 3/4 inch condulets were not i

large enough to allw splice installation without bending to radii less thanallowed. Splice installation and irspection instructions did not include bendradius specifications. j

,

Plant personnel performed extensive investigations follwing the NRC's B:)audit. On May 6,1989 utility ergineers discovered unqualified splices in Main i

Steam (SB) and Emergency Feedwater (BA) systen instrumentation wiring that wasrequired to be envii== tally qualified. During initial splice installation, i

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'work instructions for splice installation specified use of heat shrink sleeves(SIN) that were too small for the cable specified. Therefore, the installed'

splices were not envim .etally qualified. Splice installation instructions ;

were changed to specify the proper size sleeves. The documentation preparedto accomplish this change provided no nethod to assure that the inpropersplices were reworked to ocuply with the new instructions. There was also no ;

'

quality inspection plan developed to verify acceptability of the work. i|'

On May 6, 1989 utility engineers also identified unqualified cables and splicesassociated with safety related flow and level transmitters in the Reactorcoolant System (RCS) and one motor operated valve in the Makeup and i

Purification System (CB). These cables and splices were located below the'

flood level in the RB. The cables and splices were not qualified for,

sukmergence ard should not have been routed below the flood level. The 1

instructions by which the cable raceways were installed did not adequately |define all D2 requirements. )

l

IMPROm< OIL I.EVEL IN INSTRMDfr JUNCTION BJXES I

|On April 26, 1989 during the NRC D2 audit, it was discovered that the Reactor '

Rtilding Sanp level transmitters (NH,LT) and the Reactor Buildirg flood level,

transmitters had not been maintained in accordance with D2 requirements. The |

electrical junction boxes associated with the level transmitters are requiredto be filled with silicon oil to provide protection frun moisture andsubmersion. When the junction boxes were inspected they were found to haveless than the required amount of oil. This cwouaised the environmental !

'qualification of these ocuponents.

Investigators fourd no record of maintenance which would have removed the oil.Records frun the installation of the transmitters show that the transmitterswere properly filled when they were installed in 1983. However, since thattime, there has been no regular surveillance program to monitor oil level inthe junction boxes.

1

MISSING OR PAINTED OVER T-IEAINS IIRN1 AND CAPPED OR MTSSDM GREASE PT'LTEF |

FTITDGS ON VALVE M7IOR OPERATORS

On April 26, 1989 the NRC D2 audit discovered D2 deficiencies associated withfour valve motor operators located in the Reactor Buildirg. The D2deficiencies identified involved the installation ard maintnnance of motor 4

operator T-drains (enclosure drains) and grease reliefs (thermal expansion !I reliefs),

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Inspectors fourd problems associated with the followirg valves:

1) CAV-1, Pressurizer (AB,PZR) steam space sagling containment >

isolation valve (SHV),i

2) CAV-3, Pressurizer water space saglirg containment isolation valve,

3) CAV-4, Steam Generator (AB,SG) "A" sanplirg containment isolationvalve,

4

4) RCV-11, Isolation valve for Pressurizer Pilot Operated Relief Valve.

Innadiately following the NRC audit, utility personnel performed inspections ofthe 21 valve motor actuators that require envim =ud.al qualification, and arelocated in the Reactor Baildirg. 'Ibe inspections addressed installation andmaintenance in the areas of T-drains, grease reliefs, and splices andterminations associated with limit switdes. ;

'Ibe valve currently installed as RCV-11, ard ita associated motor actuator,were installed and tested in 1982. 'Ibe operator qualification test inclwiwireferences to T-drains. It should be noted that in same instances T-drains are

| shipped with motor operators, but are not attached. Similarly, grease reliefsare covered with a cap durity shipping. hwi upon current verification data,it amears that T-drains were never installed, and grease relief caps werenever removed.

Motor operators on valves CAV-1, CAV-3, and CAV-4 were replaced in 1979 due toEQ concerns. Valve operator test procedures used at that time did not includeT-drains. In 1981, plant personnel determined that the valve operators werenot qualified for subaergence, even though they were located below thepostulated flood elevation in the Reactor Building. 'Ibe valves were relocated.Relocation work did not include T-drain installation.

In 1983, valves CAV-1 and CAV-3 ard their associated operators were replaced ,

with different types of valves and operators due to operational problems.Modification instructions for installation of the new valves includeddirections for installing T-drains. However, the modification contained no

,

instructions for removing grease relief shipping caps. Based upon current '

inspection data, the T-drains were installed on CAV-1 and CAV-3 (although the '

CAV-1 T-drain was found plugged), but the grease relief shipping caps for bothCAV-1 and CAV-3 had never been removed.

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In 1985, as a result of additional reviews required by IE Bulletin 79-01B,Florida I%er Corporation (FPC) replaced 13 valve actuator notors in theReactor Baildiry with new Class RH insulation notors and pinion gears. hnodification also contained specific instructions for verifying theinstallation of T-drains and grease reliefs in 9 of the 13 actuators. Itappears the verification instructions for the 9 actuators were performedbecause the current inspection results indloated all 9 had T-drains and greasereliefs installed. However, several of these actuators had plugged greasereliefs and one had only one T-drain. From the @ mantation, it is not clearwhy the modification did not include the remaining EQ actuators located in theHaactor Building.

In 1986, plant personnel inspected and of the 21 envirewiantally qualifiedvalve actuators in the Reactor Building. 'Ihis inspection discovereddeficiencies related to T-drains and grease reliefs. h inspectioninstructions provided guidance for identifying the deficiencies and notifyingappropriate supervision. 'Ibe identified deficiencies were documented onindividual inspection data sheets whim were then forwarded to the Site NuclearProcurunent Engineer for review. It appears the ocmpleted inspection sheetsard work requests were never adequately reviewed and appropriate correctiveactions were never pursued.

ICRV Rin3 VALVE CETSICPATICH

on May 1,1989 a utility electrician found that the ICRV Block Valve controlcable insulation [ISL) and motor operator grease had deteriorated due to highambient tenperatures. 'Ibe electrician made this discovery as part of the NRCD2 audit. Valve RCV-11 is located on top of the Pressurizer. During the May 1inspection, the electrician also noted that the motor leads were not properlyspliced. 'Ibe reason for the incorrect splices appears to be that inadequateinstructions were provided when splices were installed.

During the 1981 refuelirg cutage, personnel discovered high tenperature damageto the notor control cables associated with the FORV Block Valve, RCV-11, aswell as two other valves. Plant personnel replaced the rhmgad cables andinstalled junction boxes to facilitate replacement of the cables. During the1983 refueling outage, the RCV-11 control cables were replaced again because ofheat th=ya to the insulation. During the 1985 refualing outage, plantpersonnel proposed replacement of the RCV-11 control cables with new cablesinsulated with a material resistant to high tenperature and radiation.However, the pr==3 cables would have been susceptible to damage by highhumidity. h modification was rejected, and the cables were replaced with newcables of the original type,

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%e original cables were considered to be Envim idally Qualified. However,the cable insulation will not endure long term exposure to the tanperaturesencountered in the area Where valve RCV-ll is installed. Insulation that willwithstand the tanperatures at this location are porous and may fail due to high

i humidity.,

W e RCV-ll actuator has been refurbiahed and the motor and motor control leads'

have been replaced. Also, the motor limit switch ocupartment space heater has '

been diamsi: dad. Se utility will develop a preventive maintenance program-

to inspect and replace inte-i.ect wiring as required. mis practice willcontinue unless a resolution is developed that prevents degradation of thewirirg. We RCV-ll operator motor will be replaced during the next refueling |

outage. Utility ergineers are investigating replacement of the RCV-ll motorwith a newer style motor equipped with RH insulation.

IN00RRECT PIDGS INSTAL 1ED IN PRESSURE TRANSMITTER CONDUIT 00NNECTIONS

on May 5,1989 ' tility personnel discovered plastic plugs installed in conduitucuiissctions associated with two Steam Generator pressure transmitters. We|

-

| transmitters were shipped with plastic plugs in the conduit connection( openirgs. Se plastic plugs should have been replaced with stainless eteel

plugs durirg transmitter installation. Plastic plugs remained in place due topersonnel oversight. Durirg developnent of plans and instructions for '

installing the two transmitters, personnel did not recognize the need toreplace plastic plugs with stainless steel plugs.

Plant persconel have replaced plastic plugs with stainless steel plugs. !

9UIEBOUINT ED IEFICIENCES

St*= pant to the NRC inspection and following return to IWER OPERATION on,

| | July 6,1989, additional EQ discrepancies were identified by FPC:

| At 1800 on June 30, 1989, during maintenance on Feedwater Valve 30 ( M -' 30) (SJ,V), an unqualified splice was found in the motor operator of this

valve. Se unqualified splice in M-30 was replaced with a qualifiedsplice. -

On July 7, 1989, during review of D2 rh-ntation, it was determinedthat all four channels of core flood tank level instrumentation (BP,LI)did not have proper conduit seals installed at the location where the .

conduit connects to the transmitter. A modification will be developed to ,

install conduit seals on all core flood tank level transmitters. i

Utility personnel have performed additional investigations of EQ equipment,

and EQ records during the time since EQ deficiencies were firstdiscovered mis investigation uncovered several reportable deficiencies.

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| 12MIT SWI'IGES

on September 28, 1989 utility personnel discovered nine limit switdes thatwere not erwironmentally qualified for the area in whi@ the swit&es wereinstalled. 'Ihe swit&as provided position indication for five valves in theNuclear Servloes closed Cycle Cooling Water (SW) System (CC) and four valves inthe Industrial Cboling Water (CI) System (IM). 'Ihree of the SW valves inquestion direct the flow of cooling water to the Reactor Bailding cooling Units(BK, CIR). 'Ihe other two SW valves in question isolate cooling flow to theIstdown Coolers (CC, CIR) following actuation of the Engineered Safeguards (JE)Systan. 'Ibe four CI valves isolate cooling to the space between the Reactorvessel and the Primny Shield Wall.

Unqualified swit&es were installed as a result of personnel error. Personnelresponsible for initial installation of the switches did not correctly identifythe envim==:stal conditions urder which the switches were tw3uired to operate,

h unqualified swit&es will be replaced with qualified switches by the end ofthe refuel 7 outage.

On November 30, 1989 utility ergineers discovered limit switches that weremissing cover screws. h covers were r=whi to provide environmentalprotection for the limit switches. 'Ibe problem swit6es were associated withMain Steam System valve MSV-148 [SB, SHV). Valve MSV-148 is a containmentisolation valve in the drain line frtu the "B" Steam Generator. Limit switchcover screws were ruplaced on Novenber 30, 1989.

UN00ALIFIED IIXIS AND UNIIBn'inw WIRE

On October 3, 1989 plant personnel discovrad unqualified lugs in a positionindication circuit for one of the Main Steam Isolation Valves (SB, SHV). hcircuit also contained wire that was not hwaanted as being environmentallyqualified.

Personnel could not positively identify the wire as envimw.utally qualified.

Unqualified lugs and wire were replaced with environmentally qualified lugs andwire cm October 5,1989.

UNIDENI'IFIED WIRE

On October 25, 1989 plant personnel discovered an unidentified wire in themotor starter (SA,MSIR) circuit associated with Auxiliary Steam System (SA)valve ASV-5 (SA, SHV). 'Ihis valve supplies steam to the 'Iurbine DrivenEmergency Feedwater Punp (BA, P) (EFP-2).

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| Ch Novenber 6, 1989, utility engineers discovered unqualified wires in thenotor starter circuits associated with Emergency Feedwater System valves EEV-14 ,

and EFV-33 [BA,5HV). 'Ibese two valves isolate Emergency Feedwater flow frtanthe motor driven Emeupq Feedwater Punp (BA, P).

,

i

| On Novenber 30, 1989, utility engineers discovered unidentified wires in themotor starter circuits answiated with Main Steam System valves MSV-55 and MSV-56 (SB, SHV). '1hese valves supply steam to the Turbine Driven Emergency;

Feedwater Punp.

Unidentified wires were installed due to personnel error. Iwrsonnel did notrecognize the need for h=antation of environmental qualification of wiringin the valve starter circuits.

All unidentified wires have been replaced with identified environmentally '

qualified wires. Utility personnel replaced wires associated with ASV-5 onOctober 25, 1989. Wires associated with EEV-14 and EFV-33 were replaced onNovember 6, 1989. 'Ihe MSV-55 and M3V-56 wires were replaced on Mr 1,1989.

MI

'Ihese events are varied in nature and root cause. However, the events indicate',

the overall envitsmental qualification program was deficient in the followingareas:

1) Development of overall EQ pruparu definition, responsibilities,administrative controls, and detail procedures,

2) Technical ard pra amatic training at levels or stages of programinplementation,

3) Communication and coordination of program requirements andresponsibilities ery to achieve and maintain desired programobjectives,

4) Post EQ installation verification, inspection ard acceptance,

| 5) Maintenance of EQ performance capability; i.e., specific EQsurveillance prugas ard procedures, specific EQ preventativemaintenance activities.

6) Insufficient controls to assure corrective actions related to designdeficiencies are inplemented.

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;

1) BEND RADII IESS 'IHAN ALIDWED BY MhNUFACIURER'S GUIDEIJNESi

A) Affacted T4 =rit : RC-3A-PT3, RC-3B-PT3, [AB, PT) l!iRC-14A-DPT1, RC-14A-DPT2, DC14A-DPT3, RC-14A-DPT4, RC-14B-DPT1,

RC-14B-DPT2, RC-14EFDPT4 [AB, FT)

| Transmitters RC-3A-PT3 and RC-3B-PT3 monitor RCS pressure. 'Ihey jprwide input signals for actuation of the Engineered SafeguardsSysten (ES) [JE). 'Ihe other transmitters listed above prwide RCSflw signals to the Reactor Protection Syctan [JC) (RPS) for theFlux /Flw Imbalance Trip.

Failure of these transmitters would not have occurred unless a harsh Ienvim__ it existed in the Reactor Building. Such conditions would !only exist follwing a loss of Coolant Accident (IOCA) or Main Steam I

line break in the RB. If either of these events occurred, ES and/orRPS actuation should nmm before splice failure occurred. Also, theEngineered Safegur2ds and Reactor Protection Systans monitor other i

| parameters that would cause the systens to actuate. *

I'Ihe berxl radius lower limit of five diameters was based on the boundsof the splice manufacturer's analysis. 'Ihe manufacturer had no data :

to verify whether or not splices would function properly with berri '

radii less than five diameters. Florida Power Corporation obtainedtest reports concernirxJ this issue fran another utility. 'Ihe testreports indicate that the type of splice in question would maintain ,

its qualification at bend radii of one diameter or less. 'Iberefore,there is lw probability that these splices would have failed due totheir bend radius. 'Ihe splices in question may be considered to bequalifiable. |

2) SPLICE SIEEVES 'ICO SMALL, INCORRECT PIDGS INSTALLED IN PRESSURE +

TRANSMTITERS

iAffected F? != ant (Splice Sleeves): MS-106-PT, MS-107-PT, MS-108-PT,A) ,

MS-109-PT, MS-110-PT, MS-111-PT, MS-112-PT, and MS-113-PT [BA, PT) '

Affected Equipnent (Inchoct Plugs): MS-111-PT and MS-113-PT*

'Ihese instruments sense pressure in the secondary side of the SteamGenerators (SG) and transmit signals to the Emertjency FeedwaterInitiation and Control System (EFIC) [BA). Signals frun thesei

;

instruments are used for Main Steam and Feedwater isolation,initiation of Emergency Feedwater (EFW) flw to the Steam Generators, ;

and control of feedwater f1w.

| knC Form 3 A 1649),

_ _ _ , . . - _ . . -_ - ___ _ . _

. - .. . . . _ _ . - . . _,

!PORu ambA U.S. NUCLE As s tOULX,ORY C0eAMt98604 '

EKPlIt8 of3002!

. UCENSEE EVENT REPORT (LERI ,'?'s"^'!,9Auji'|',P,}'J'ya"5!,'0,,c?T',' &*' '.O,

TEXT CONTINUATION J"!$ ,'o'/ilt';0"cj &W(* lf|c',", ^||2 '"' MS"u :

n?W,'a,"JJi","il = a',M"te?"d4*Mi,'Ci' '

i- ;

OF MANAGEMENT AND SUDGtT W A36tlNGTON.DC 20603

| F ACILITY h4Mt nl DDCKlf N(*MBER (28 ggg gygggR (Sp PAOI (31 |'

, , , , , a o . ~, .. , . . . .CRYSTAL RIVER UNIT 3 """ """

,

0|3 1|0 or 1|5q 1| 60 |5 |0 |0 |0 ] 3|0 p 8 |9 - -

irixta . e ct assawnn '

Failure of three or more of these instruments on either SG in an '

energency situation requiring EFW, would prevent proper EFW actuation| or Steam Generator . isolation, or could cause urwarranted EFW

actuation or Steam Generator isolation. Failure of these instruments '

could also prevent proper EFW flow rate durirq Natural Circulation. ,

In either of these events, operators would be able to manuallyoperate mydmant to isolate SG's, unisolate SG's, or control EFW '

flow.

l A harsh envitu. it in the Intermediate Ih111 ding would result from a,

Main Steam or Main Feedwater line Break in the building. In theevent of a Main Steam or Main Feedwater Line break, EFIC shouldisolate Main Steam and Main Feedwater bsfore splices failed due toharsh envitu i nt. Splice failure later in either event coulddefeat the EFIC logic that controls EFW flow during RCS naturalcirculation flow, or the logic that prevents EFW flow to a faultedSteam Generator. If EFIC did not autmatically isolate Main Steamand Main Feedwater, operators would be able to manually isolate thesei

systens.

Florida Power Corporation has obtained data fr m another utility that !

d= w .^uates that the splices in question were qualifiable.

B) Affected nyiimant: EF-24-PP, EF-25-Pf, and EF-26-PP (BA, PT) [

'Ibese instruments measure Emergency Feedwater flow, and provide flowindication to the Main Control Board. Tma of these instrumentswould not directly prevent proper control of Emergency Feedwater i

flow. However, lost or failed indication could mislead operatorsduring a transient.

,

Florida Power Corporation has obtained data frun another utility tP4 t^

.

dsig uates that the splices in question were qualifiable.

3) SUBERSION

A) Affected Eeyiimant RC-14A-DPT1, RC-14A-DPr2, RC-14A-DFT3, ;

RC-14B-DPfl, RC-14B-DPr2, ard RC-14B-DPf3 (AB, PI) ~

Each of these instruments provide signals to the Reactor ProtectionSystem for the Flux / Flow / Imbalance 'frip. Also, transmittersRC-14A-DPfl, RC-14A-DPf2, RC-14B-DPr1 ard RC-14B-DPI2 pIUvideirdication of Reactor Coolant Systenn flow on the Main Control Board.

4

NRC Fe 3B&A M91._. _ _ . , _ . . _ - ,

alltC f onm ae6A U.S. NUCLE Am b t ULI TOR Y COMMlessoNAPPRovt0 0MS NO. 3164It04

6EX787.t$ 4/30/92

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0?"t#a"M i",",a*R M = ie?"a*# & 0?iCf ;08 MANAGE Mt B(T AND DVDGtT. yr AsmaNGToN, DC 20(4J

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F ACILif y esamt its DOCILLY NUMetR (21 ggg gygggg ggi pagg g3) !

CRYSTAL RIVER UNIT 3 I

q 1| 6 Oj3 1|1 or 1| 5 jo p Io lo |o | 3|0 2 8|9 _

*TEXT ,J mese apsee e essesset, was asisumpur 44C #wm mW (17)

Envitu et.al qualification of these instruments could be umptudseddue to RB flooding concerns. Reactor Building floodirg sufficient tothreaten operability of these transmitters would only nmm followinga Icos of Coolant Accident. We major source of RB floodirg ;

following a IJXA would occur as a result of water injected by the ;

Engineered Safeguards systesn. Se RPS will trip the reactor beforeES actuation since the RPS setpoints are higher than or equal toEngineered Safeguards setpoints. We subject transmitters are notrequired to function after a reactor trip. % e same reasoning would;

apply in the case of a small break IDCA too small to initiate ES.'

Tbliowing a IDCA which depleted the entire RCS volume, there would beno RCS flow to monitor. Werefore the transmitters need not be i

functional following a large break IOCA. Indication of RC flowfollowing a small break IOCA that did not deplete the entire RCSvolume could be derived fran the status of other RCS parameters. -

|

B) Affected Egiimant: RC-1-LT3 (AB, LT]

Wis transmitter provides the signal used for automatic Pressurizerlevel control, as well as indication of Pressurizer level on the Main '

control Board.

Following a IDCA or Main Steam or Main Feedwater line Break, RCSinventory could be controlled by the High Pressure Injection (HPI)(BJ) and low Pressure Injection (LPI) (BP] systems. loss of thecontrol signal fran RC-1-LT3 at this point would not hinder transient '

mitigation. However, this transmitter is required by NuclearRegulatory Guide 1.97 to be functional for a period of 24 hoursfollowing either a large or small break IOCA.

C) Affected Egi4mant: RC-3A-PI4 (AB, PT)

Wis instrument provides an RCS pressure signal to initiate ES.Since the major source of RB flooding is the ES system, the safety '

function of this instrument would be aoocmplished before floodingnm wred. Wis transmitter is also required by Nuclear RegulatoryGuide 1.97 to mitigate the consequences of a IDCA.

s

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WMC pere 3e6A (649)

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. LICENSEE EVENT REPORT (LER)$",*,'#,o',$8WeWJ'no"ji,'%'N# N.".'J'sTEXT CONTINUATION R"",'yo'."4 M'"0"d *f"f'AUleCl# l,".'"'ENf!-

"'P'tNa".'J "*a* A **!"e?M*a*d* M ?ici'

Of MANAGE Mt NT AND SUDGET,W ASHINGTON. DC 20603.

, Ac,tiTY hAME tu Dt *ti NUMBl. (2) Lth NUR.Dth 1 1 PAOS (31

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CRYSTAL RIVER UNIT 3 """" * " *

0 |5 |0 |0 |0 | 3| 0 |2 8|9 Oll|6 0|3 1| 2 0F ll 5- -

.Exv ,x e m w. .as w we s amu w nn

D) Affected Equipment: SP-21-LT, SP-22-LT, SP-23-LT, SP-24-LT,SP-31-LT, and SP-32-LT [AB, LT)

These transmitters monitor level in the "B" Steam Generator.Transmitters SP-21-LT, SP-22-LT, SP-23-LT and SP-24-LT monitor "HighRange" level. Transmitters SP-31-LT and SP-32-LT monitor "Iow Range"level. These instruments prwide a signal for EFIC actuation, andEIM Block Valve (BA,SHV) ocotrol. These transmitters are requiredfor the proper operation of the Emergency Feedwater System. Thetransmitters are required to operate 24 hours post accident.

The EFIC system controls EIM flow based on "High Rarge" levelindication during RCS natural cirullation flow. The syst e also uses |

"High Range" level indication for initiation of Steam Generatoroverfill protection. Failure of the SP-21-LT or SP-22-LT would causeimproper EfW flow control during RCS natural' circulation flow.Failure of SP-23-LT or SP-24-LT would cause EFW Block Valves to closeprematurely, or would prevent valves fran closi%. when required. Ineither case, operators would be able to manually control E!W flow oroperate EIW Block Valves as tw = == y.

Failure of the " low Rarge" instruments could cause premature EFICactuation, or prevent EFIC actuation on two of the four EFICchannels. In the event of a LOCA, Engineered Safeguards systemactuation would actuate the EFIC system independently of SteamGenerator levels before RB flooding occurred. In the event of a MainSteam or Main Feedwater Line Break, EFIC system would actuate due tolow Steam Generator Pressure. Therefore, Reactor Building floodingdue to these events would not prevent EFIC actuation. Since both ofthese events would require EFIC actuation, Reactor Building floodingwould not cause pra ature actuation due to failure of thesetransmitters.

E) Affected Equipnent: RC-163A-LT1, RC-163B-LT1, FC'-164A-LTl and RC-164B-LTl (AB, LT)

The transmitters provide Reactor Vwel (AB,RPV) water levelindication to the Reactor Coolant Inventory Tracking System (RCITS)This system is not required for transient mitigation. Failure ofthese instruments would not degrade performance of ES equipment orhinder accident mitigation capabilities.

es.Cfeem 3.SA (EHl.l

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0|3 1|3 or 1| 5q lj 6o [s |0 |o |o | 3|0 |2 8|9 '- .-

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| F) Affected M it: Valve MN-505 Motor Operator (CD,10]

Valve MN-505 (CB,lSO) is a containment isolation valve for one ofthe three letdown coolers (CB, HX]. Se valve is closed by ES

i

actuation, and is required to remain closed. In the event of a IOCAand mooanpanying ES actuation, the valve will perfom its isolationfunction before RB flooding occurs. Once closed, the valve couldnot reopen if the operator were fludswi.

I he valve position indication limit switches would short out if I

fludsd, and position indication would be lost to the Main ControlBoard. mis would not de p ade performance of ES equipnent or hinder-

accident mitigation capabilities.

IMHOPER OIL IEVEL IN INS'IWJMElff JUNC1'ICN BOXES

j Affected Equiptent: WD-301A-LT, WD-301B-LT, WD-302A-LT, WD-302B-LT, WD-303A-LT,WD-303B-LT, WD-304A-LT, WD-304B-LT

Transmitters WD-30lVB-LT and WD-302VB-LT are the Reactor Building Sunp level |

transmitters. Transmitters WD-303VB-LT ard WD-304VB-LT measure ReactorBuilding Flood Isvel (water level above the Reactor Building Floor). %ese

,

transmitters provide indication on the main control board. Se instruments )provide no autanatic control function. Ioss of indication frm these |instruments would not prevent operation of ES equipnent. However, operators |

use indications from the flood level instrunents when swapping IPI Punp (BP,P] |,

'

suction frut the Borated Water Storage Tank (BP,TK] to the Reactor Building )sunp during a IDCA. Loss of indication frun the flood level transmitters would

locmplicate the transition. I

During a IDCA, the sunp level transmittats may fail due to submersion.| However, as water level would continue to rise, the flood level instnments

would provide adequate level indication. Se flood level transmitters arelocated above the Reactor Building flood level. Werefore, it is not likely -

that the flood level instruments would fail due to flooding. I

CAPPED OR MISSING T-NAINS AND VENIS ON VALVE M7IOR OPERA'IORS

Valves CAV-1, CAV-3, and CAV-4 autanatically close upon receipt of anautanatic diverse containment isolation signal frun the ES system. 'Ihesevalves pruiptly receive an ES signal to close and will have performed their i

safety function before being exposed to a harsh envirunment. Each of thesevalves have redundant containment isolation valves outside of the RB. In theevent of IDCA, the outboard valves would still be available for containmentisolation.

i

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NRC ,sein astA (649)1

. . . _ _ . . ._ , _ _ . _

g,.N. A v . oct.Aa._um,0., C- N,,,,, ,,, _ ,, ,,,,

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. UCENSEE EVENT REPORT (LER) f,$||"J7|,%',"ig'*/cP,J'5%"ji,*.|yTV T"J"'!,

TEXT CONTINUATION *j|",',",'o' "!"Od u"?infle'#'/M '"' O!!"N,

1 P Attav 0 pt itDN R l'3 60 De O .COF MANAGEMENT AND DuDGET,UgA.MINGTON.DC 70 03.

# ACILt1T NAM 4113 DOCK 41 NUMDE R (26 gg g ,gggg a ggi pagg g3i

T''" waY.we *CRYSTAL RIVER UNIT 3 -

0|5|o|o|o|3|0 p 8 |9 0j 1| 6 0 |3 1|4 ,oF 1|5- -

78KT fa nose apsee 4 fuepeat use esWenno' MIC , ann 354W tih

P:RV BIDCK VALVE IE11!RIORATIQ4

If RCV-11 failed in the closed ocruliticri, there is no safety sigraficance. IfBCV-11 and the PCRV both failed open, cperators would not be able to isolateflow thrux$ the PCRV. However, operators would be able to maintain RCSinventory via the HPI and IPI systens. Plant small break IDCA analysis boundsthis event.

In the event of a IDCA and w'1ying low RCS pressure, the N would notopen autcznatically. Operators would have no reason to open the valve manually.'Iberefore, it is not likely the N and RCV-ll would both fail open in thisevent. If the RCS repressurized following a IDCA, operators would be requiredto use the IWV to maintain RaB pressure below 2300 psig. In this scenario,there is a possibility that the IWV arri PORV block valve would both fail open.

suBSIEuBNF ED IEFICIENCIES

FWV-30 is the main block valve in the feedwater ficw path to Steam Generator"B" (AB,SG). '!he safety function of this valve is to close on a low steamgenerator pressure actuation of the Emergency Feedwater Initiation and ControlSystem (EFIC) (BA). 'Ihere are other valves in the flow path which also closeon this signal which would also isolate this flow path.

'Ihe core flood tank level instrumentation is used by operators to mordtor arx1maintain the core flood tanks at the proper level during normal operations.Following an accident, these indicators are used to verify proper operation ofthe Core Flood Tanks (BP,TK). 'Ihis instrumentation is qualified for its normalenvironment. 'Ibe accident in which the core flood tanks are r=iel to ensurethat the ECCS acceptance criteria are met is the large break IOCA. In thistype of accident, the Core Flood tanks enpty within the first few minutes afterthe break occurs. Once the tanks have enptied and the operators have verifiedthat they are enpty, the core _ flood tank level instrumentation is no longernaariad. Due to the short time frame during which operation in an accidentenvironment is required, FFC determined that the level instrumentation wasoperable.

LIMIT SWrIG ES

'Ibe limit switches provide position indication only.

UNQUALIFmn IDGS AND UNIDEknnw WIRE

'Ihe circuit containirx3 the unqualified owents provided position indicationonly. Failure of the lug or wire would affect irxlication only. Valveoperation would zw ain unaffected.

i

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a, UCENSEE EVENT REPORT (LER) ,',0|",,^4'?,4%8?"si'of,'SUdT%CET.!/ To'.".1"2'

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I AC8LITY NA184 06 DoctLif WuMDth 421 tlk WUtept. 2) PA.4 (3)

.... . ovi . 6 ..v* * " " " " " " "

CRYSTAL RIVER UNIT 3

d Il 6 013 11 5 0F 1h0 ]5 |0 | 0 | 0 | 31012 819 - -

Text in - . W.c % m.w nn

UNIMNTIFIED WIRE

Failure .of any of the unidentified wires may have prevented actuation of theaffected valves. Valve ASV-5 is normally maintained closed. If the valvefailed in the closed position, motive steam to EFP-2 would still be available

,

via Auxiliary steam Valve ASV-204. Valves MSV-55 and MSV-56 are normally'

maintained open. If the valves failed open, steam: supply to EPP-2 would be.

unaffected. Because valves MSV-55 and MSV-56 are not located in the same area !'as valve ASV-5, it is not likely that all three valves would fail open due to

harsh envitu d. at the same time. Therefore, if any one of these valvesfailed open due to harsh envitu. dal conditions, operators would be able toisolate the steam supply to EFP-2. In the unlikely event that all three of thevalves failed open, operators would still be able to stop EFP-2 by tripping the ,

punp.

Valves EFV-14 and EFV-33 are normally maintained open. If the valves failed inthis position, operators would be able to isolate flow frun the Motor Driven :EFW Punp by tripping the punp, or by closing the EFW control valves. i

.

CGWBCTIVE ACTIG( |

| iIn order to prevent future nmmrences, the utility has cannitted to perform EQ '

training in August 1989 and to inplement an enhamsisit to the present EQprogram. 'Ihe enhancement will address the following seven areas: '

a. Organizationb. Proceduresc. Field Verificationd. Documentation :e. Envitu.iedal Profilef. EQ Master List ig. Trainirg ;

i

PREVIOUS SIMIIAR EVENFS,

The utility has subnitted five previous Licensee Event Reports concerning !

envitu.-Aal qualification deficiencies.

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'macF.= a.na nee_. , . . . . . . . . _ _ _ _ _ _ _ _