ML19327B230
| ML19327B230 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/23/1989 |
| From: | Moffatt L, Widell R FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F1089-12, LER-89-016, LER-89-16, NUDOCS 8910270215 | |
| Download: ML19327B230 (14) | |
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C OR POR ATION October 23, 1989 3F1089-12 .j U. S. Nuclear Regulatory Cunnission Attention:' Document control Desk Washington, D. C.-20555
Subject:
Crystal River Unit 3 .;
Docket No. 50-302 Operating License No. DPR-72 ,
, Licensee Event Report No. 89-016-02
Dear Sir:
Enclosed is supplement two for Licensee Event Report (IER)89-016 which is submitted in accordance with 10 CPR 50.73. An additional supplement 7 to address any items identified during the currunt Envium u-d.al Qualification walkdown outside of containment will be subnitted by ,
January 15, 1990.
Should there be any questions, please contact this office. ;
Very truly yours, b <
RolfiC. Widell Director, Nuclear Operations Site Support +
WIR: mag Enclosure ,
e xc: Regional Administrator, Region II Senior Resident Inspector i
8910270210 891023 PhR
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UCENSEE EVENT REPORT (LER) EIMMiN r'T*o*No'i'O fU""'EYlL#ET* 0IN "'Ful .?tCc'RM AND HEPORTS MANAGEME NT ORANCH (P4301. U $ NUCLE AR RE GULATORY COMMIS$10N. W ASHINGTON DC 70666. AND TO THE PAPIRWOHK Rf DUCTION PROJECT t316041041 Of f lCE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20bO3.
PACILITY 8sAME m DOC K E T NUMSE R (21 FAGE (31 CMSTAL RIVER UNIT 3 0 l 5 l 0 l 0 l 0 l3 l0 l2 1joFl1l3 ,
Administrative _ Problems Caused Deficiencies in the Environmental Qualification Program Resulting in Plant Equipment Not Properly Qualified IVGNT DATE ill LE R NUMet h (Si REPORT DAf t Ifl OTHE R F ACILITill $NVOLVED (St MONTH DAY YtAR YEAR OM L'6 7,5,$ MONTH DAv YEAR 8 Acstet. N Avts DOCKET NUMetRtBI ,
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j Crystal River Unit 3 was in IODE 5 (ODID SHUIDOWN) frun February 27, 1989 to June 1, 1989. During this outage, NRC inspectors discovered deficiencies related to envimuminal qualification of plant equipnent. Deficiencies
- included improper cables and splices, improper silicon oil level in instrurcent
, junction boxes, and problems related to valve motor operators. Problems were l the result of deficiencies in detailed development and implementation of the l- environmental qualification program. Utility personnel have repaired l
identified envimumuial qualification deficiencies, or have justified continued operation with the deficiencies until repairs are ccupleted. 'Ihe l utility has embarked on a major voluntary effort to review the existing l Environmental Qualification program, ard to correct additional environmental qualification deficiencies that may be discovered.
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EVENP IESCRIPI'IQ4 Crystal River Unit 3 was in PODE 5 (COID SHUIDOWN) frm February 27, 1989 to June 1, 1989. Durirg this cutage, a Nuclear Regulatory Comnission inspection team discovered several deffciencies related to environmental qualification (D2) of Plant Equipnent. Inspectors found the following deficiencies:
1
- 1) Impmper electrical cable (CBL) aM splice (CON) installation, ,
including cables and splices not qualified for subnersion found located below the Reactor Building (RB)(NH) flood level,
- 2) Inproper oil level in instrument junction boxes [JBX), ,
1
- 3) Missirg or painted over T-drains (IEN], and miscing or capped grease l relief fittings on valve notor operators (84), l
- 4) Detericration of wires and grease associated with the Pilot Operated Relief Valve (10RV) (AB,RN] Block Valve (Tag No. RCV-11) (AB,SHV).
UNOUALIFIED CABI.ES ICBL1 AND SPLTCES ICON 1 On April 26, 1989, during the NRC Environmental Qualification (D2) FWuuu audit, inspections fouM that cable splices on signal cables frm two pressure transmitters (PT) had not been installed in accordance with the splice ,
l manufacturer's application guide. We application guide required that each l splice beM radius be no smaller than five times the outside splice diameter.
! 2e manufacturer had no data to determine whether or not splices could be
- i. qualified with smaller bend radii. Inspectors found bend radil that were only l two to three times the outside radii of the splices. %ese splices were located in the Reactor Coolant System (AB) instrumentation wiring, between conduit seal assemblies and the field cables.
I Original plans called for installation of junction boxes between the instrument conduit seal assemblies and the field conduits. Wese boxes were to be large enough to allow splice installation with acceptable bend radii. Due to seismic mounting difficulties, the plans were revised to specify 3/4 inch condulets (CDT) instead of the junction boxes. W e 3/4 inch condulets were not large enough to allcw splice installation without bending to radii less than I allowed. Splica installation and inspection instructions did not include bend radius specifications.
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0 l1 l6 0 l2 0l3 or 1 l3 Textu - . o .we w miaim Plant personnel performod extensive investigations following the NRC's B2 audit. On May 6,1989 utility engineers discovered ungt.nlified splices in Main Steam (SB) and Emergency Feedwater (BA) system instrumentation wiring that was required to be environmentally qualified. During initial splice installation, work instructions for splice installation specified use of heat shrink sleeves (SLV) that were too small for the cable specified. Therefore, the installed splices were not environmentally qualified. Splice installation instructions were changed to specify the proper size sleeves. The rhunmtation prepared to aoxmplish thas change provided no method to assure that the ihprwer splices were reworked to conply with the new instructions. There was also no quality inspection plan developed to verify acceptability of the work.
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 aM Purification System [CB). These cables aM splices were located below the flood level in the RB. The cables and splices were not qualified for submergence and should not have been routed below the flood level. The instructions by which the cable raceways were installed did not adequately define all D2 requirements.
IMPROPER OIL I.EVEL IN INSTRUMENT JUNCTION BOXES On April 26, 1989 during the NRC EQ audit, it was discovered that the Reactor l Building Sump level transmitters (NH,LT] aM the Reactor Building flood level I
transmitters had not been maintained in accordance with EQ requirements. The I
electrical junction boxes associated with the level transmitters are required
! to be filled with silicon oil to provide protection from moisture' and I submersion. When the junction boxes were inspected they were found to have less than the required amount of oil. This cumnised the environmental qualification of these cmponents.
Investigators found no record of maintenance which would have removed the oil.
l Records frcm the installation of the transmitters show that the transmitters l 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.
MISSING CR PAINTED OVER T-DRAINS TDRN1 AND CAPPED OR MISSING GREASE PFLTEF FITTINGS ON VALVE MOIOR OPERATORS I
l On April 26, 1989 the NRC EQ audit discovered EQ deficiencies associated with four valve motor operators located in the Reactor Building. The EQ deficiencies identified involved the installation aM maintenance of motor operator T-drains (enclosure drains) and grease reliefs (thermal expansion l
l reliefs).
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0l2 0 l4 oF 1p i rixtin . u. ,* wcr., o,nn Inspectors found problens aaam iated with the following valves:
- 1) GV-1, Pressurizer (AB,PZR] steam space sagling contaiment l' isolation valve (SIN), j l l CAV-3, Pressurizer water space sagling containment isolation valve,
- 2) j J
- 3) CAV-4, Steam Generator (AB,SG) "A" sa@ ling containment isolation l I
valve,
- 4) RCV-11, Isolation valve for Pressurizer Pilot Operated Relief Valve.
Idiately following the NRC audit, utility personnel performed inspections of the 21 valve motor actuators that require environmental qualification, and are located in the Reactor Etilding. 'Ibe inspections addr==1 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, and its associated motor actuator, '
were installed and tested in 1982, h operator qualification test included references to T-drains. It should be noted that in sane instances T-drains are shipped with motor operators, but are not attached. Similarly, grease reliefs i are covered with a cap during shipping. Based upon current verification data, it appears that T-drains were never installed, and grease relief caps were never removed.
s Motor operators on valves CAV-1, CAV-3, and GV-4 were replaced in 1979 due to D2 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 stimidgence, even though they were located below the postulated flood elevation in the Reactor Building. h valves were relocated.
Relocation work did not include T-drain installation.
In 1983, valves CAV-1 ard CAV-3 and 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.
NRC 7 sem .e.A (6419)
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0 l2 0l5 OF 1 l3 text <n e . n. .M.-=Ac i amm on l In 1985, as a result of additiorel reviews required by IE Bulletin 79-OlB, Florida Power Corporation (FPC) replaced 13 valve actuator motors in the Raactor Building with new Class IE insulation motors aM pinion gears. %e modification 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 indicated all 9 had T-drains and grease reliefs installed. However, several of these actuators had plugged grease reliefs aM one had only one T-drain. From the documentation, it is not clear why the mdification did not include the remaining EQ actuators located in the Reactor Building.
]
! In 1986, plant personnel inspected each of the 21 environmentally qualified '
valve actuators in the Reactor Building. Wis inspection discovered l deficiencies related to T-drains and grease reliefs. We inspection I instructions provided guidance for identifying the deficiencies and notifying l appropriate supervision. Se identified deficiencies were dmananted on I individual inspection data sheets which were then forwarded to the Site Nuclear i Procui.wun. Engineer for review. It appears the ccmpleted inspection sheets I and work requests were never adequately reviewed and appropriate corrective actions were never pursued.
ICRV BIDCK VALVE DETERIORATION l
On May 1,1989 a utility electrician found that the FORV Block Valve control cable insulation [ISL] and motor operator grease had deteriorated due to high ambient tenperatures. We electrician made this discovery as part of the NRC EQ 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. Se reason for the incorrect splices appears to be that inad=mte instructions were provided when splices were installed.
During the 1981 refueling outage, personnel discovered high tenperature damage to the motor control cables associated with the IORV Block Valve, RCV-11, as well as two other valves. Plant personnel replaced the damaged 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 damage to the insulation. Durirg the 1985 refueling outage, plant personnel proposed replacement of the RCV-11 control cables with new cables insulated with a material resistant to high temperature and radiation.
However, the proposed cables would have been susceptible to damage by high humidity. We modification was rejectal, and the cables were replaced with new cables of the original type.
NRC Foren asea (6491
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'Ihe original cables were considemd to be Enviwiumirally Qualified. However, the cable insulation will not endure long term exposure to the tenperatures encountered in the area Aw.re valve RCV-ll is installed. Insulation that will withstand the tenperatures at this location are porous and may fail due to high humidity.
'Ibe RCV-11 actuator has been refurbished and the motor and motor control leads have been replaced. Also, the motor limit switti capartment space heater has been removed. 'Ihe utility will develop a preventive maintenanoa program to inspect and replace intewuivisct wiring as required, and will replace the RCV-11 operator motor during the next refueling outage. Utility emineers are investigating replacement of the RCV-11 motor with a newer style motor equipped with RH insulation. Such a motor would be rated for higher tenperatures than j the currently installed motor. 1 1
INwKKtd PI1X;S INSTAT.Tm IN PRESSURE TRANSMITTER OJNIUIT CONNECTIONS l l
On May 5,1989 utility personnel discovered plastic plugs installed in conduit connections associated with two steam Generator pressure transmitters. 'Ibe transmitters were shipped with plastic plugs in the conduit connection openings. 'Ihe plastic plugs should have been replaced with stainless steel plugs during transmitter installation. Plastic plugs runained in place due to ,
personnel oversight. Durirg development of plans and instructions for
- installing the two transmitters, personnel did not recognize the need to replace plastic plugs with stainless steel plugs.
Plant personnel have replaced plastic plugs with stainless steel plugs.
SUBSEQUENP ED DEFICIENCIES Sph=pt to the NRC inspection and followirg return to IMER OPERATION on July 6, 1989, two additional EQ discrepancies were identified by FPC:
At 1800 on June 30, 1989, during maintenance on Feedwater Valve 30 (FW-
- 30) (S7,V), an unqualified splice was found in the motor operator of this valve, j On July 7, 1989, during review of D2 documentation, it was determined I that all four channels of core flood tank level instrumentation (BP,LI]
I did not have proper conduit seals installed at the location where the l conduit connects to the transmitter.
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'1hese events are varied in natum arx1 root cause. However, the events irx11cate the overall einironmental qualification program was deficient in the follwing areas:
- 1) Developnent of overall EQ program definition, responsibilities, ,
administrative controls, and detail procedures,
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1
- 2) 'Dachnical arx1 prv mTiriatic training at levels or stages of program iglementation,
- 3) Communication and coordination of program requirements and res.ruwibilities ecey to achieve ard maintain desired pwpan objectives,
- 4) Post m installation verification, inspection and acceptance,
- 5) Maintenance of M perfomance capability; i.e., specific EQ surveillance programs ard procedures, specific EQ preventative maintenance activities.
- 6) Insufficient controls to assure corrective actions related to design deficiencies are iglemented.
EVENT ANALYSIS IMPROPER PTPCIRTCAL CARTR AND SPLTCE INSTATI ATION, INCORRECT PIDGS INSTATJRn IN l PRESSURE 'mAMMPPPER ONDRT CONNECPINS 1
- 1) BEND RADII I.ESS 'IHAN ALIDWED BY MANUFACIURER'S GUIDELINES l A) Affected Equipment : RC-3A-Pf3, RC-3B-PT3, [AB, Pr]
I RC-14A-DFT1, RC-14A-DFT2, RC14A-DPf3, RC-14A-DFI4, RC-14B-DPfl, RC-14B-DPf2, RC-14B-DPr4 [AB, FT]
Transmitters RC-3A-PT3 and RC-3A-PI3 monitor RCS pressure. 'Ihey provide input signals for actuation of the Engineered Safeguards System (ES) [JE] . 'Ibe other transmitters listed above provide RCS flow signals to the Reactor Prutection System [JC] (RPS) for the Flux / Flow Imbalance Trip.
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Failure of these transmitters would not have occurmd unless a hamh envis.-ud. existed in the Ranctor M1dirig. Sucts canditions would only exist.followirg a Ioss of coolant Accident (INA) or Main Steam l line break in the RB. If either of theee events occurred, ES and/or RPS actuation abould occur before splice failum rmered. Also, the Engineered Safeguards and Reactor Protection Systems monitor other parameters that would cause the systems to actuate.
L, i
'Ibe band radius lower limit of five diameten was bas x1 on the bourxSs of the splice manufacturer's analysis. 'Ibe manufacturer had no data to verify whether or not splices would fWrtion properly with bend i radii less than five diameters. Florida FNer Corporation obtained test reports concerning this issue frun amther utility. 'Ihe test reports irxiicate that the type of splice in question would maintain its qualification at bend radii of one dianuter or 1Ns. 'Iberefore, v there is low probability that these splices would have failed due to their bend radius. 'Ihe splices in qtustion may be considerod' to be qualifiable. .
- 2) SPIJCE SIEEVES 'IOO - SMALL, IN00RRECT PHES INSTArim IN PRESSURE
'IRANSMTPTERS L ,
A) Affected Equipnent (Splice Sleeves): MS-106-PI, MS-107-PP, E -108-PT, MS-109-PT, 2 -110-PT, MS-111-PT, E -112-PT, and MS-113-PI (BA, PT]
Affected Equipment (Incorrect Pltgs): MS-111-PT and MS-113-PT p 'Ibese instruments sense pressure in the secondary side of the Steam
, Generators (SG) and transmit signals to the Emergency 'Feedwater l Initiation and Control System (EFIC) [BA). Signals from these i
instruments are used for Main Steam and Feedwater isolation, initiation of Emergency Feedwater (EFW) flow to the Steam Generators, and control of feedwater flow.
Failure of three or more of these instruments on either SG in an ,
emergency situation requirirg EFW, would prevent proper EFW actuation or Steam Cenerator isolation, or could cause unwarranted EFW actuation or Steam Generator isolation. Failure of these instruments could also prevent proper EFW flow rate durirg Natural Circulation.
In either of these events, operators would be able to manually operate equipnent to isolate SG's, unicolate SG's, or control EFW flow.
NRC Form 306A (6 89)
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- 1.xv u . , = w =c P 3siaw nn A harsh envim.isit in the Inte=diate Building would result frun a i 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 before splices failed due to l harsh envirw.isit. Spljoe failure later in either event could o defeat the EFIC logic that controls EIW flow during RCS natural circulation flow, or the logic that prevents EIW f1w to a. faulted
- . Steam Generator. . If EFIC did not automatically isolate Main Steam l and Main Feedwater, operators would be able to manually isolate these systems, t I l
. Florida Power Corporation has obtained data fran another utility that '
desid=Lcates that the splices in question were qualifiable.
l B) Affected Equipment: EF-24-FT, EF-25-FT, and EF-26-FT [BA, FT)
! 'Ibese instruments measure Emergency Feedwater flow, and provide flow indication to the Main Control Board. Loss of these instruments would not directly prevent proper control of Emergency Feedwater l i
flow. However, lost or failed indication could mislead operators during a transient.
Florida Power Corporation has obtained data frun ancfdwr utility that dimis=Ltates that the splices in question were qualifiable.
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- 3) SUINERSION l A) Affected Equipnent: RC-14A-DFF1, RC-14A-DFT2, RC-14A-DFT3, RC-14B-DFT1, RC-14B-DPI2, and RC-14B-DFT3 (AB, FT]
Each of these instruments provide signals to the Reactor Protection System for the Flux / Flow / Imbalance Trip. Also, transmitters l RC-14A-DFF1, RC-14A-DFT2, RC-14B-DFT1 and RC-14B-DPT2 provide indication of Reactor Coolant System flow on the Main Control Board. I Environmental qualification of these instruments could be wytmised j due to RB flooding concerns. Reactor Building flooding sufficient to !
threaten operability of these transmitters would only occur following l a Loss of Coolant Accident. 'Ihe major source of RB flooding i following a IOCA would occur as a result of water injected by tle ,
Engineered Safeguards system. 'Ihe RPS will trip the reactor before ES actuation since the RPS setpoints are higher than or equal to Etgineered safeguards setpoints. 'Ibe subject transmitters are not required to function after a reactor trip. 'Ihe same reasonirg would apply in the case of a small break IOCA too small to initiate ES.
r Nic 7erm 3.SA (689)
tegic P0ltu MSA V4 NucLER kBoutata.Y C0edMIGRicer APPRoyED oms No.31604104 M .- EKPlIll 4f30/IP2 -
LICENSEE EVENT REPORT (LER) lSs*L',%'n80,y,",of;',';o"li,'%cf','T ,T.".!"n2 I "
TEXT CONTINUATlON ?"",f,",'o'.'i!",17%%it.f"fi%We'##1,In".TETN L
Mit.",'a"JJt"fa'#M' Eta *4t*& t"? 'ci L Of MANAGEMENT AND SUDGET W ASHINGTON.DC 20603.
9AC4LITY 9s4844 (O DOCR&T NUMSE R (2) - ggg hyggggly (g) PAag (3) ,
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Qyeaa CRYSTAL RIVER UNIT 3 0l5l0l0l0l3l0l2 8 l9 ._
0 l1 l6 0l2 1l0 or 1l3 1sut w -, .= = =,.w mwn.w me r== answ nn Follwing a IOCA which depleted the entire RCS volume, there wculd be no RCS flw to nonitor. 'Iherefore the transmitters need not be
r B) Affected Equipment: RC-1-LT3 [AB, E)
'Ihis transmitter provides the signal used for autanatic Pressurizer level control, as well as indication of Pressurizer level on the Main control Board.
Following a IOCA or Main Steam or Main Feedwater line Break, RCS inventory could -be controlled by the High Pressure Injection (HPI)
[B7) and Iow Pressure Injection (IPI) [BP) systems. Loss of the control signal frun 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 IDCA. -(
1 C) Affected Equipment: RC-3A-Pr4 [AB, PF) j l 'Ihis instrument provides an RCS pressure signal to initiate ES.
l Since the major source of RB flooding is the ES system, the safety
! function of this instrument would be acccuplished before flooding occurred. 'Ihis transmitter is also required by Nuclear Regulatory Guide 1.97 to mitigate the consequences of a IOCA.
D) Affected Equipment: SP-21-LT, SP-22-LT, SP-23-LT, SP-24-III', -
l SP-31-LT, arrl SP-32-LT [AB, LT) l l 'Ihese transmitters monitor level in the "B" Steam Generator.
! Transmitters SP .11-E, SP-22-2, SP-23-LT and SP-24-LT monitor "High l
Range" level. Transmitters SP-31-LT and SP-32-LT monitor "Iow Range" i level. 'Ihese instruments provide a signal for EFIC actuation, and EFW Block Valve [BA,SHV) concrol. 'Ihese transmitters are required for the proper operation of the Dtergency Feedwater System. 'Ihe ;
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.
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'Ibe EFIC system controls EFW flow based on "High Range" level indication during RCS natural circulation flow. 'Ibe system also uses "High Range" level fix31 cation for initiation of Steam Generator overfill protection. Failure of the SP-21-IIP or SP-22-E would cause improper EFW flow control during RCS natural circulation flow.
Failure of SP-23-LT or SP-24-IIP would cause EFW Block Valves to close prematurely, or would prevent valves from closing when required. In either case, operators would be able to manually control EFW flow or operate EFW Block Valves as necessary.
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- j, LICENSEE EVENT REPORT (LER)- s*E',%',fMEdWo$'!3"MT%CfWl
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,T."W.".'s M ?s!'.ML'C7"f,'!E 7,".' t'efdA TEXT CONTINUATION 1 P A,t RWO Rtt J TION (3 to 04 1 OF MANAGlutNT AND OUDGET, WASHINGTON,0C 30903.
FACILITY 8 eat #E tu DUCILEi NUM8tR (33 LtR NUhASSR tt) PA06 (3)
~ CRYSTAL RIVER UNIT 3-0 l6 j o' l 0 l 0 l 3 l0 l 2 8l9 -
0 l1 l6 -
0l2 1- l 1 or 1l3 vuxt en ., ,. e wane, asnoinn <
. Failure of the "Im Rarge" instruments could cause prenature EFIC '
actuation, or prevent EFIC actuation on two of the four EFIC channels. In the event of a IDCA, Engineered Safeguards system actuation would actuate the EFIC systen irdspanimitly of Steam Generator levels before RB flooding occurred. In the event of a Main Steam or Main Feedwater Line Break, EFIC systeen would actuate due to lcw Steam Generator Pressure. 'Iberefore, Reactor Buildirg flooding due to these events would not prevent EPIC actuation. Since both of these events would require EFIC actuation, Reactor Building flooding would not cause premature actuation due to failure of these
- transmitters.
[ E) Affected Equipnent: RC-163A-LT1, RC-163B-LT1, RC-164A-LT1 and RC-164B-LT1 [AB, LT] ;
'Ihe transmitters provide Reactor Vessel [AB,RPV) water level indication to the Reactor Coolant Inventory Tracking System (RCITS) .
'Ihis system is not required for transient mitigation. Failure of i these instruments would not degrade performance of ES equipnent or hinder accident mitigation capabilities.
F) Affected Equipaent: Valve 70V-505 Motor Operator (CB, M0] f Valve MUV-505 (CB, ISO] is a containment isolation valve for one of the three letdown coolers [CB, FDC) . 'Ihe valve is closed by ES actuation, ard is mquired to remain closed. In the event of a IDCA and arrv==mying ES actuation, the valve will perform its isolation furction before R3 flooding occurs. Once closed, the valve could ;
not reopen if the operator were flevviarl.
'Ihe valve position indication limit switches would short out if !
flooded, and position irdication would be lost to the Main Control ,
Board. 'Ihis would r.ot degrade performance of ES equipnent or hindez.
accident mitigation capabilities.
IMPROPER OIL IEVEL IN INSTRUMENI' JUNCI' ION B0igis Affected Equipmerrt: WD-301A-LT, WD-301B-LT, WD-302A-LT, WD-302B-LT, WD-303A-III',
WD-303B-IIP, WD-304A-LT, WD-304B-LT Transmitters WD-301A/B-IIP ard WD-302A/B-IIP are the Reactor Building Sunp level transmitters. Transmitters WD-303A/B-IIP ard WD-304A/B-LT measure Reactor Building Flood Invel (water level above the Reactor Building Floor). 'Ihese transmitters provide indication on the main control board. 'Ihe instruments provide no automatic control function. Ioss of indication frun these instruments would not prevent operation of ES equipment. However, operators NRC f eem 3e6A (6491
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I FORM 196A U.S.5:UCLEAR f.E AULATOxY COMMI.860N , g t kPlats 4/30/92
' *" WS"00'PoMTl ,,oCf"W IE.".1"'!
lFM#,0N UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION C,$"4",'o','l',' M21"oiU"?'.MflCU,135 L"'lMf"fA O!"tAla"M t"M* A *,.' E !!? W o'k Ofic?
OF MANAGEMENT AND BUO!1tT WASHINGTON. DC 70603, 1 F ACaLITV Naut 01 DOCKET NUMBER (21 Lin NUMeth 141 PA06 (31
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CRYSTAL RIVER UNIT 3 I o l5 lo lo lo l 3l 0l2 8l 9 ._
0 l1 l 6 ._
0 l2 1l2 OF 1 l3 rixt <n - % M , , ac r.,,,, su.m use indications fmn the flood level instruments when swapping IPI Punp (BP,P) I suction frm the Borated Water Storage Tank (BP,TK] to the Reactor Building sunp during a IDCA. Loss of indication from the flood level transmitters would ccurplicate the transition.
During a LOCA, the sunp Invel transmitters may fall due to subnersion. '
However, as water level would continue to rise, the flood level instmments would provide adequate level indicction. %e flood level transmittem am located above the Reactor Building flood level. Werefore, it is not likely that the flood level instruments would fail due to flooding.
CAPPFD OR MISSING T-CRAllC AND VENTS ON VALVE M7IOR OPERA'IORS Valves CAV-1, CAV-3, and CAV-4 autcutically close upon receipt of an autcanatic diverse containment isolation signal frm the ES system. %ese valves pmtptly raceive an ES signal to close aM will have performed their safety function befom being exposed to a harsh environment. 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.
PORV BLOCK VALVE DETERIORATION L If RCV-11 failed in the closed condition, there is no safety significance. If RCV-11 and the ICRV both failed open, operators would not be able to isolate flow through the PORV. However, operators would be able to maintain RCS inventory via the HPI and LPI systems. Plant small break IDCA analysis bounds i this event.
1 l In the event of a IDCA and aNnying low RCS pressure, the PORV would not open autcmtically. Operators would have no reason to open the valve manually.
Berefore, it is not likely the PORV aM RCV-11 would both fail open in this event. If the RCS repressurized following a IOCA, operators would be required to use the ICRV to maintain RCS pressure below 2300 psig. In this socnario, there is a possibility that the FORV and FORV block valve would both fail open.
SUBSIIKMFI' B0 DEFICIENCIES l FWV-30 is the main block valve in the feedwater flow path to Steam Generator "B" (AB,SG]. Se 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). Bere are other valves in the f1w path which also close on this signal which would also isolate this flow path.
E".C Form 346A 16891
PO8tM 306A U.S.NUCLEt.2 R51ULf TOWY COMMeOS404 p ikPIIES 4/30/02 e
LICENSEE EVENT REPORT (LER) ISg.t'A o'o u"8'" Ma "5,'Po"v'!,'
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0F MANAGtMENT AND SUDGET VW A&HING1ON, DC 20603.
, PACILITV NAMt H) DOCEti NUMBER Qi LER NUMSIR 446 PA06 (31 CRYSTAL RIVER UNIT 3 o p jo jo jol3l0 l 2 8l 9 -
0 l1 l6 -
0 l2 1l3 0F 1 l3 Text n - . w. -ac % .mu w nn me core flood tank level instrumentation is used by operators to monitor and 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 instrumentaticn is qualified for its normal envisumant. h accident in which the core flood tanks are needed to ensure that the EOCS acceptance criteria are met is the large break IMA. In this type of accident, the Core Flood tanks empty within the first few minutes after the break occurs. Once the tanks have aiptied and the operators have verified that they are enpty, the core flood tank level instrumentation is no longer i W. Due to the short time frame during which operation in an accident ,
environment is required, FIC determined that the level instrumentation was j operable.
CORRECTIVE ACTIN l
'Ibe deficiencies which were identified in the inspection were corrected during the 1989 spring outage. 'Ihe unqualified splice in FW-30 has been replaced ,
with a qualified splice. A modification will be developed to install conduit seals on all core flood tank level transmitters. l In order to prevent future occurrences, the utility has cmmitted to perform EQ training in August 1969 and to inplement an enhancement to the present EQ program. N enhancement will address the following seven areas: ,
- a. Organization
- b. Procedures
- c. Field Verification l d. h w ntation
- e. Enviiumental Profile l f. EQ Master List I
- g. Training PREVIOUS SIMIIAR EVDfIS
'Ihe utility has submitted five previous Licensee Event Reports concerning environmental qualification deficiencies.
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