ML19354D909
| ML19354D909 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/17/1990 |
| From: | Moffatt L FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19354D908 | List: |
| References | |
| LER-89-016, LER-89-16, NUDOCS 9001240380 | |
| Download: ML19354D909 (15) | |
Text
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"' ADMINISTRATIVE PROBLEMS CAUSED DEFICIENCIES IN THE ENVIRONMENTAL QUALIFICATIOb P ROG RAM RESULTING IN PLANT EQUIPMENT NOT PROPERLY QUALIFTED.
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NO l l l AwuAC,,4,, io --.. ,,,... ..A, ,. .i..,nei Crystal River Unit 3 was in PODE 5 (CDI.D SHUI'DOWN) fran February 27, 1989 to June 1, 1989. Durirg this outage, NRC inspectors discove ed deficiencies related to envirw isital qualification of plant equipnant. S e p nt investigations of envite.ustA qualification records have discovered additional deficiencies. Deficiencies included improper cables and splices, inproper silicon oil level in instrument junction boxes, and problems related to valve motor operators. Problens were the result of deficiencies in detailed developnent and inplementation of the envite..sital qualification program.
Utility penminal have repaired identified envire..ard qualification deficiencies, or have justified continued operation with the deficiencies until repairs are ocupleted. 'Ihe utility has enbarked on a major voluntary effort to review the existirq Envile. td Qualification propath, and to correct additional envim isital qualification deficiencies that may be discovered.
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0l3 0l2 or 1l5 l Text an . v. un wme r mawmi i EVIMP MEKRIPPIG( j Crystal River Unit 3 was in MXE 5 (CDID SHlTfDOWN) frun February 27, 1989 to l June 1, 1989. Ikaring this outage, a Nuclear Regulatory Ocamission inspection l l team disoevered several deficiencies related to environmental qualification l (BQ) of Plant hWnt. Inspectors fourd the follwing deficiencies: ;
l l 1) Igroper electrical cable (CBL) and splice (CON) installation, !
including cables and splices not qualified for subnersion found I located belw the Reactor Bailding (RB)(NH) flood level, !
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- 2) Igrger oil level in instrument junction boxes (JHX),
- 3) Missing or painted over T-drains (mW), and missing or capped grease relief fittings on valve motor operators (84),
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- 4) Deterioration of wires and grease ==wiated with the Pilot Operated j Relief Valve (PORV) (AB,RN) Block Valve (Tag No. RCV-ll) [AB,SHV). )
i UNOUALIFTEn CAmm fCBL1 AND SPLTCES f 00N) l l On April 26, 1989, during the NRC Envilu =uiuti Qualification (EQ) Program l audit, inspections found that cable splices on signal cables frun two pressure I transmitters (PT) had not been installed in accordance with the splice manufacturer's application guide. 'Ihe application guide required that each splice 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 be qualified with smaller bend radii. Inspectors found bend radii that were only two to three times the outside radii of the splices. 'Ihese splices wero i located in the Reactor Coolant System (AB) instrumentation Wiring, between conduit 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 inch coMulets (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 than allowed. Splice installation and irspection instructions did not include bend radius 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 was required to be envii== tally qualified. During initial splice installation, i i
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ttKT IF mow apose 4 #seuser. see esluttene r N4C Ace 35541(th 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 prepared to accomplish this change provided no nethod to assure that the inproper splices were reworked to ocuply with the new instructions. There was also no ;
quality inspection plan developed to verify acceptability of the work. i l
On May 6, 1989 utility engineers also identified unqualified cables and splices associated with safety related flow and level transmitters in the Reactor coolant 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 l define all D2 requirements. )
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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 l electrical junction boxes associated with the level transmitters are required to be filled with silicon oil to provide protection frun moisture and submersion. When the junction boxes were inspected they were found to have less 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 transmitters were properly filled when they were installed in 1983. However, since that time, there has been no regular surveillance program to monitor oil level in the junction boxes.
1 MISSING OR PAINTED OVER T-IEAINS IIRN1 AND CAPPED OR MTSSDM GREASE PT'LTEF l FTITDGS ON VALVE M7IOR OPERATORS On April 26, 1989 the NRC D2 audit discovered D2 deficiencies associated with four valve motor operators located in the Reactor Buildirg. The D2 deficiencies identified involved the installation ard maintnnance of motor 4 operator T-drains (enclosure drains) and grease reliefs (thermal expansion !
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Taxv<n ==.. e.i ean-w wea 3simtm Inspectors fourd problems associated with the followirg valves:
- 1) CAV-1, Pressurizer (AB,PZR) steam space sagling containment >
isolation valve (SHV),
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- 2) CAV-3, Pressurizer water space saglirg containment isolation valve,
- 3) CAV-4, Steam Generator (AB,SG) "A" sanplirg containment isolation valve, 4
- 4) RCV-11, Isolation valve for Pressurizer Pilot Operated Relief Valve.
Innadiately following the NRC audit, utility personnel performed inspections of the 21 valve motor actuators that require envim =ud.al qualification, and are located in the Reactor Baildirg. 'Ibe inspections addressed installation and maintenance in the areas of T-drains, grease reliefs, and splices and terminations 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 inclwiwi references to T-drains. It should be noted that in same instances T-drains are l shipped with motor operators, but are not attached. Similarly, grease reliefs are covered with a cap durity shipping. hwi upon current verification data, it amears that T-drains were never installed, and grease relief caps were never removed.
Motor operators on valves CAV-1, CAV-3, and CAV-4 were replaced in 1979 due to EQ concerns. Valve operator test procedures used at that time did not include T-drains. In 1981, plant personnel determined that the valve operators were not qualified for subaergence, even though they were located below the postulated 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 included directions 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 both CAV-1 and CAV-3 had never been removed.
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0l1l6 - q3 q5 0F 1 l5 TEXT (2 mee ausse 4 seewest, amo esweener Nec Femm W (176 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 the Reactor Baildiry with new Class RH insulation notors and pinion gears. h nodification also contained specific instructions for verifying the installation of T-drains and grease reliefs in 9 of the 13 actuators. It appears the verification instructions for the 9 actuators were performed because the current inspection results indloated all 9 had T-drains and grease reliefs installed. However, several of these actuators had plugged grease reliefs and one had only one T-drain. From the @ mantation, it is not clear why the modification did not include the remaining EQ actuators located in the Haactor Building.
In 1986, plant personnel inspected and of the 21 envirewiantally qualified valve actuators in the Reactor Building. 'Ihis inspection discovered deficiencies related to T-drains and grease reliefs. h inspection instructions provided guidance for identifying the deficiencies and notifying appropriate supervision. 'Ibe identified deficiencies were documented on individual inspection data sheets whim were then forwarded to the Site Nuclear Procurunent Engineer for review. It appears the ocmpleted inspection sheets ard work requests were never adequately reviewed and appropriate corrective actions were never pursued.
ICRV Rin3 VALVE CETSICPATICH on May 1,1989 a utility electrician found that the ICRV Block Valve control cable insulation [ISL) and motor operator grease had deteriorated due to high ambient tenperatures. 'Ibe electrician made this discovery as part of the NRC D2 audit. Valve RCV-11 is located on top of the Pressurizer. During the May 1 inspection, the electrician also noted that the motor leads were not properly spliced. 'Ibe reason for the incorrect splices appears to be that inadequate instructions were provided when splices were installed.
During the 1981 refuelirg cutage, personnel discovered high tenperature damage to the notor control cables associated with the FORV Block Valve, RCV-11, as well as two other valves. Plant personnel replaced the rhmgad cables and installed junction boxes to facilitate replacement of the cables. During the 1983 refueling outage, the RCV-11 control cables were replaced again because of heat th=ya to the insulation. During the 1985 refualing outage, plant personnel proposed replacement of the RCV-11 control cables with new cables insulated with a material resistant to high tenperature and radiation.
However, the pr==3 cables would have been susceptible to damage by high humidity. h modification was rejected, and the cables were replaced with new cables of the original type, t%CFenn3e64(6491
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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 tanperatures encountered in the area Where valve RCV-ll is installed. Insulation that will withstand 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 will continue unless a resolution is developed that prevents degradation of the wirirg. We RCV-ll operator motor will be replaced during the next refueling l outage. Utility ergineers are investigating replacement of the RCV-ll motor with a newer style motor equipped with RH insulation.
IN00RRECT PIDGS INSTAL 1ED IN PRESSURE TRANSMITTER CONDUIT 00NNECTIONS on May 5,1989u ' tility personnel discovered plastic plugs installed in conduit l cuiissctions associated with two Steam Generator pressure transmitters. We l 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 to personnel oversight. Durirg developnent of plans and instructions for '
installing the two transmitters, personnel did not recognize the need to replace 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 l l July 6,1989, additional EQ discrepancies were identified by FPC:
l 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 qualified splice. -
On July 7, 1989, during review of D2 rh-ntation, it was determined that 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 first discovered mis investigation uncovered several reportable deficiencies.
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0l3 0l7 oF 1l5 l 12MIT SWI'IGES on September 28, 1989 utility personnel discovered nine limit switdes that were not erwironmentally qualified for the area in whi@ the swit&es were installed. 'Ihe swit&as provided position indication for five valves in the Nuclear Servloes closed Cycle Cooling Water (SW) System (CC) and four valves in the Industrial Cboling Water (CI) System (IM). 'Ihree of the SW valves in question 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 the Istdown Coolers (CC, CIR) following actuation of the Engineered Safeguards (JE)
Systan. 'Ibe four CI valves isolate cooling to the space between the Reactor vessel and the Primny Shield Wall.
Unqualified swit&es were installed as a result of personnel error. Personnel responsible for initial installation of the switches did not correctly identify the envim==:stal conditions urder which the switches were tw3uired to operate, h unqualified swit&es will be replaced with qualified switches by the end of the refuel 7 outage.
On November 30, 1989 utility ergineers discovered limit switches that were missing cover screws. h covers were r=whi to provide environmental protection for the limit switches. 'Ibe problem swit6es were associated with Main Steam System valve MSV-148 [SB, SHV). Valve MSV-148 is a containment isolation valve in the drain line frtu the "B" Steam Generator. Limit switch cover screws were ruplaced on Novenber 30, 1989.
UN00ALIFIED IIXIS AND UNIIBn'inw WIRE On October 3, 1989 plant personnel discovrad unqualified lugs in a position indication circuit for one of the Main Steam Isolation Valves (SB, SHV). h circuit also contained wire that was not hwaanted as being environmentally qualified.
Personnel could not positively identify the wire as envimw.utally qualified.
Unqualified lugs and wire were replaced with environmentally qualified lugs and wire cm October 5,1989.
UNIDENI'IFIED WIRE On October 25, 1989 plant personnel discovered an unidentified wire in the motor starter (SA,MSIR) circuit associated with Auxiliary Steam System (SA) valve ASV-5 (SA, SHV). 'Ihis valve supplies steam to the 'Iurbine Driven Emergency Feedwater Punp (BA, P) (EFP-2).
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. rext <* === == = <. .asa w we inn, answ im f l Ch Novenber 6, 1989, utility engineers discovered unqualified wires in the notor starter circuits associated with Emergency Feedwater System valves EEV-14 ,
and EFV-33 [BA,5HV). 'Ibese two valves isolate Emergency Feedwater flow frtan the motor driven Emeupq Feedwater Punp (BA, P). ,
i l On Novenber 30, 1989, utility engineers discovered unidentified wires in the motor starter circuits answiated with Main Steam System valves MSV-55 and MSV-
Unidentified wires were installed due to personnel error. Iwrsonnel did not recognize the need for h=antation of environmental qualification of wiring in the valve starter circuits.
All unidentified wires have been replaced with identified environmentally '
qualified wires. Utility personnel replaced wires associated with ASV-5 on October 25, 1989. Wires associated with EEV-14 and EFV-33 were replaced on November 6, 1989. 'Ihe MSV-55 and M3V-56 wires were replaced on Mr 1, 1989.
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'Ihese events are varied in nature and root cause. However, the events indicate ,
the overall envitsmental qualification program was deficient in the following areas:
- 1) Development of overall EQ pruparu definition, responsibilities, administrative controls, and detail procedures,
- 2) Technical ard pra amatic training at levels or stages of program inplementation,
- 3) Communication and coordination of program requirements and responsibilities ery to achieve and maintain desired program objectives,
- 4) Post EQ installation verification, inspection ard acceptance, l 5) Maintenance of EQ performance capability; i.e., specific EQ surveillance prugas ard procedures, specific EQ preventative maintenance activities.
- 6) Insufficient controls to assure corrective actions related to design deficiencies are inplemented.
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0l3 1l 5 rut ,,, -. w. u c r mar.nm EVENT ANhI2 SIS IMROPER EIECIRICAL CABLE AND SPLICE INSTALLATIQL INCCRRECT PIDGS INSIALIED IN l PRESSURE 'IRANSMTffER 03EUIT CONNBCTIONS
- 1) BEND RADII IESS 'IHAN ALIDWED BY MhNUFACIURER'S GUIDEIJNES i
A) Affacted T4 =rit ! : RC-3A-PT3, RC-3B-PT3, [AB, PT) l RC-14A-DPT1, RC-14A-DPT2, DC14A-DPT3, RC-14A-DPT4, RC-14B-DPT1, i RC-14B-DPT2, RC-14EFDPT4 [AB, FT) l Transmitters RC-3A-PT3 and RC-3B-PT3 monitor RCS pressure. 'Ihey j prwide input signals for actuation of the Engineered Safeguards Systen (ES) [JE). 'Ihe other transmitters listed above prwide RCS flw signals to the Reactor Protection Syctan [JC) (RPS) for the Flux /Flw Imbalance Trip.
Failure of these transmitters would not have occurred unless a harsh I envim__ 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/or RPS actuation should nmm before splice failure occurred. Also, the Engineered Safegur2ds and Reactor Protection Systans monitor other i l parameters that would cause the systens to actuate.
- I
'Ihe berxl radius lower limit of five diameters was based on the bounds of 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 obtained test reports concernirxJ this issue fran another utility. 'Ihe test reports 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 to their bend radius. 'Ihe splices in question may be considered to be qualifiable.
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- 2) SPLICE SIEEVES 'ICO SMALL, INCORRECT PIDGS INSTALLED IN PRESSURE +
TRANSMTITERS A) Affected F? !=i ant (Splice Sleeves): MS-106-PT, MS-107-PT, MS-108-PT, ,
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 Steam Generators (SG) and transmit signals to the Emertjency Feedwater i
Initiation and Control System (EFIC) [BA). Signals frun these ;
instruments are used for Main Steam and Feedwater isolation, initiation of Emergency Feedwater (EFW) flw to the Steam Generators, ;
and control of feedwater f1w.
l knC Form 3 A 1649) ,
PORu ambA U.S. NUCLE As s tOULX,ORY C0eAMt98604 '
EKPlIt8 of3002 UCENSEE EVENT REPORT (LERI ,'?'s"^'!,9Auji'l',P,}'J'ya"5!,'0,,c?T',' &*' '.O TEXT CONTINUATION J"!$ ,'o'/ilt';0"cj &W(* lflc',", ^ll2 u '"' MS :
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OF MANAGEMENT AND SUDGtT W A36tlNGTON.DC 20603 l F ACILITY h4Mt nl DDCKlf N(*MBER (28 ggg gygggR (Sp PAOI (31 l
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CRYSTAL RIVER UNIT 3 """ """
i 0 l5 l0 l0 l0 ] 3l0 p 8 l9 -
q 1l 6 -
0l3 1l0 or 1l5 rixta . 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 l 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 manually operate 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 the event of a Main Steam or Main Feedwater Line break, EFIC should isolate Main Steam and Main Feedwater bsfore splices failed due to harsh envitu i nt. Splice failure later in either event could defeat the EFIC logic that controls EFW flow during RCS natural circulation flow, or the logic that prevents EFW flow to a faulted Steam Generator. If EFIC did not autmatically isolate Main Steam i and Main Feedwater, operators would be able to manually isolate these 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 flow indication to the Main Control Board. Tma of these instruments would not directly prevent proper control of Emergency Feedwater i flow. However, lost or failed indication could mislead operators during a transient. ,
Florida Power Corporation has obtained data frun another utility tP4 t .
dsig uates that the splices in question were qualifiable.
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- 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 Protection System for the Flux / Flow / Imbalance 'frip. Also, transmitters RC-14A-DPfl, RC-14A-DPf2, RC-14B-DPr1 ard RC-14B-DPI2 pIUvide irdication 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 COMMlessoN APPRovt0 0MS NO. 3164It04 EX787.t$ 4/30/92 6 U
. LICENSEE EVENT REPORT (LER) lUf1Rdo*N M"l'cWol'M8"fni .o*ff,'l ,7.".1"! l TEXT CONTINUATlON E"".'6",'o','?,*Mff!"MMI," !"c'"^lMU".' fi'E7f! i 0?"t#a"M i",",a*R M = ie?"a*# & 0?iCf ;
08 MANAGE Mt B(T AND DVDGtT. yr AsmaNGToN, DC 20(4J l
F ACILif y esamt its DOCILLY NUMetR (21 ggg gygggg ggi pagg g3) !
CRYSTAL RIVER UNIT 3 I o p Io lo lo l 3l0 2 8l9 _
q 1l 6 Oj3 1l1 or 1l 5 j TEXT ,J mese apsee e essesset, was asisumpur 44C #wm mW (17)
Envitu et.al qualification of these instruments could be umptudsed due to RB flooding concerns. Reactor Building floodirg sufficient to threaten operability of these transmitters would only nmm following a 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 before ES actuation since the RPS setpoints are higher than or equal to Engineered Safeguards setpoints. We subject transmitters are not required 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 be no RCS flow to monitor. Werefore the transmitters need not be i functional following a large break IOCA. Indication of RC flow following a small break IOCA that did not deplete the entire RCS volume could be derived fran the status of other RCS parameters. -
l B) Affected Egiimant: RC-1-LT3 (AB, LT]
Wis transmitter provides the signal used for automatic Pressurizer level control, as well as indication of Pressurizer level on the Main control Board.
Following a IDCA or Main Steam or Main Feedwater line Break, RCS inventory could be controlled by the High Pressure Injection (HPI)
(BJ) and low Pressure Injection (LPI) (BP] systems. loss of the control signal fran RC-1-LT3 at this point would not hinder transient '
mitigation. However, this transmitter is required by Nuclear Regulatory Guide 1.97 to be functional for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following 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 flooding nm wred. Wis transmitter is also required by Nuclear Regulatory Guide 1.97 to mitigate the consequences of a IDCA.
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I WMC pere 3e6A (649)
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EKPiltS ef3W92
. LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION $",*,'#,o',$8WeWJ'no"ji,'%'N#
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"'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 vtan st ovt hn A6 mt v isio%
CRYSTAL RIVER UNIT 3 """" * " *
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Olll6 -
0l3 1l 2 0F ll 5 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 "High Range" level. Transmitters SP-31-LT and SP-32-LT monitor "Iow Range" level. These instruments prwide a signal for EFIC actuation, and EIM Block Valve (BA,SHV) ocotrol. These transmitters are required for the proper operation of the Emergency Feedwater System. The transmitters are required to operate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident.
The EFIC system controls EIM flow based on "High Rarge" level indication during RCS natural cirullation flow. The syst e also uses l "High Range" level indication for initiation of Steam Generator overfill protection. Failure of the SP-21-LT or SP-22-LT would cause improper EfW flow control during RCS natural' circulation flow.
Failure of SP-23-LT or SP-24-LT would cause EFW Block Valves to close prematurely, or would prevent valves fran closi%. when required. In either case, operators would be able to manually control E!W flow or operate EIW Block Valves as tw = == y.
Failure of the " low Rarge" instruments could cause premature EFIC actuation, or prevent EFIC actuation on two of the four EFIC channels. In the event of a LOCA, Engineered Safeguards system actuation would actuate the EFIC system independently of Steam Generator levels before RB flooding occurred. In the event of a Main Steam or Main Feedwater Line Break, EFIC system would actuate due to low Steam Generator Pressure. Therefore, Reactor Building flooding due to these events would not prevent EFIC actuation. Since both of these events would require EFIC actuation, Reactor Building flooding would not cause pra ature actuation due to failure of these transmitters.
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 level indication to the Reactor Coolant Inventory Tracking System (RCITS)
This system is not required for transient mitigation. Failure of these instruments would not degrade performance of ES equipment or hinder accident mitigation capabilities.
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. LICENSEE EVENT CEPORT (LER) '$%,
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CRYSTAL RIVER UNIT 3 'N[YE o [s l0 lo lo l 3l0 l2 8l9 - q lj 6 .-
0l3 1l3 or 1l 5
- tan in - . w. we % mu w nn l F) Affected M it
- Valve MN-505 Motor Operator (CD,10]
Valve MN-505 (CB,lSO) is a containment isolation valve for one of i
the three letdown coolers (CB, HX]. Se valve is closed by ES actuation, and is required to remain closed. In the event of a IOCA and mooanpanying ES actuation, the valve will perfom its isolation function before RB flooding occurs. Once closed, the valve could not 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 Control Board. 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 l transmitters. Transmitters WD-303VB-LT ard WD-304VB-LT measure Reactor Building 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 l instruments would not prevent operation of ES equipnent. However, operators use indications from the flood level instrunents when swapping IPI Punp (BP,P] l 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 l ocmplicate the transition. I During a IDCA, the sunp level transmittats may fail due to submersion.
l However, as water level would continue to rise, the flood level instnments would provide adequate level indication. Se flood level transmitters are located 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 an autanatic diverse containment isolation signal frun the ES system. 'Ihese valves pruiptly receive an ES signal to close and will have performed their i safety function before being exposed to a harsh envirunment. Each of these valves have redundant containment isolation valves outside of the RB. In the event of IDCA, the outboard valves would still be available for containment isolation.
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E NRC ,sein astA (649) 1
g,.N. A v. oct.Aa._um,0., C- N IK? LIE $'Of3092
. UCENSEE EVENT REPORT (LER) f,$ll"J7l,%',"ig'*/cP,J'5%"ji,*.lyTV T"J"'!
TEXT CONTINUATION *jl",',",'o' "!"Od u"?infle'#'/M '"' O!!"N 1 P Attav 0 pt itDN R l'3 60 De O .C OF MANAGEMENT AND DuDGET,UgA.MINGTON.DC 70 03.
- ACILt1T NAM 4113 DOCK 41 NUMDE R (26 gg g ,gggg a ggi pagg g3i T" we
- waY.
CRYSTAL RIVER UNIT 3 -
0l5lololol3l0 p 8 l9 -
0j 1l 6 -
0 l3 1l4 ,oF 1l5 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. If BCV-11 and the PCRV both failed open, cperators would not be able to isolate flow thrux$ the PCRV. However, operators would be able to maintain RCS inventory via the HPI and IPI systens. Plant small break IDCA analysis bounds this event.
In the event of a IDCA and w'1ying low RCS pressure, the N would not open 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 this event. If the RCS repressurized following a IDCA, operators would be required to 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 steam generator pressure actuation of the Emergency Feedwater Initiation and Control System (EFIC) (BA). 'Ihere are other valves in the flow path which also close on this signal which would also isolate this flow path.
'Ihe core flood tank level instrumentation is used by operators to mordtor arx1 maintain the core flood tanks at the proper level during normal operations.
Following an accident, these indicators are used to verify proper operation of the Core Flood Tanks (BP,TK). 'Ihis instrumentation is qualified for its normal environment. 'Ibe accident in which the core flood tanks are r=iel to ensure that the ECCS acceptance criteria are met is the large break IOCA. In this type of accident, the Core Flood tanks enpty within the first few minutes after the break occurs. Once the tanks have enptied and the operators have verified that they are enpty, the core _ flood tank level instrumentation is no longer naariad. Due to the short time frame during which operation in an accident environment is required, FFC determined that the level instrumentation was operable.
LIMIT SWrIG ES
'Ibe limit switches provide position indication only.
UNQUALIFmn IDGS AND UNIDEknnw WIRE
'Ihe circuit containirx3 the unqualified owents provided position indication only. Failure of the lug or wire would affect irxlication only. Valve operation would zw ain unaffected.
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a, UCENSEE EVENT REPORT (LER) ,',0l",,^4'?,4%8?"si'of,'SUdT%CET.!/ To'.".1"2 TEXT CONTINUATlON Su ""4"4*."'. .^2d 'EN'." Ul#7.'!J,f 5".' L',E "Ai 0 ? W N.'Ji'7 & a",o*^3 J 2i' "id! M M?ili OF MAhAGEM!.NT AND $uDGET WASHINGTON,0C 20603.
I AC8LITY NA184 06 DoctLif WuMDth 421 tlk WUtept. 2) PA.4 (3)
.... . ovi . 6 ..v CRYSTAL RIVER UNIT 3 0 ]5 l0 l 0 l 0 l 31012 d Il 6 11 5 0F 1h 819 013 Text in - . W.c % m.w nn UNIMNTIFIED WIRE Failure .of any of the unidentified wires may have prevented actuation of the affected valves. Valve ASV-5 is normally maintained closed. If the valve failed 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 valves failed open due to harsh envitu. dal conditions, operators would be able to isolate the steam supply to EFP-2. In the unlikely event that all three of the valves 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 in this 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( l l i In order to prevent future nmmrences, the utility has cannitted to perform EQ '
training in August 1989 and to inplement an enhamsisit to the present EQ '
program. 'Ihe enhancement will address the following seven areas:
- a. Organization
- b. Procedures
- c. Field Verification
- d. Documentation :
- e. Envitu.iedal Profile
- f. EQ Master List i
- g. Trainirg ;
i PREVIOUS SIMIIAR EVENFS ,
The utility has subnitted five previous Licensee Event Reports concerning !
envitu.-Aal qualification deficiencies.
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