ML20005D662

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LER 89-026-00:on 891106,inadvertent Primary Containment Isolation Actuation Occurred While Clearing out-of-svc. Caused by Inadequate Logic Setup During Mod Installation. Trip Status & Output Switches repositioned.W/891206 Ltr
ML20005D662
Person / Time
Site: LaSalle Constellation icon.png
Issue date: 12/06/1989
From: Crowl D, Diederich G
COMMONWEALTH EDISON CO.
To:
Office of Nuclear Reactor Regulation
References
LER-89-026-01, LER-89-26-1, NUDOCS 8912140051
Download: ML20005D662 (11)
v • d • e


Text

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

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Comm:nwealth Edison q , LaSalle County Nuclear Station .

j. Rural Route #1, Box 220 -

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~

. Marseilles, Illinois 61341

. Telephone 815/357-6761 1

(j i

i December 6, 1989 I

1 Director of Nuclear Reactor Regulation U.S.-Nuclear Regulatory Commission.

Mail Station P1-137 Washington, D.C. 20555

Dear. Sir:

Licensee Event Report #89-026-00, Docket 9050-373 is being submitted to your office in accordance with

Wb G. J. Diederich h(stationManager LaSalle County Station OJD/DAC/kg '

Enclosure .

xc: Nuclear Licensing Administrator NRC Resid6c..t Inspector NRC Region III Administrator IMPO - Records Center m

e ,1 m e e 1 e1m

[DR ADOCK 05000373

- #,n PDC

. ~_. _ -. .. _ -. . - - - - - . _ _ - -

s LICENSEE EVENT REPORT (LER) p g Docket Neber (2) Pane (3)

Facility Name (I) 1 01 51 01 01 01 31 71 3 1 of 1 0

% 11e County Station unit i Title (4) Inadvertent Primary Containment Isolation Actuation Due to inadequate yakSetuoDurinoModificationInstallation 7;n-et Date (7) Other Facilities involved (8)

Event Date (5) LER Number (6)

Month Day Year Facility Names Docket N r ,er(s)

Sequential Revision Montt Day Year Year //

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/ Neber

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/// Neber 01 51 01 01 01 l i 81 9

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01 0 112 01 6 81 9 01 51 01 01 01 l l 11 1 01 6 81 9 HI U D NM TO THE REAMEMS M M OPERATING (Chock one or .e of the followine) (11) 20.405(c) ,, X,_ 50.73(a)(2)(lv) 73.71(b) 4 20.402(b) ,,_,_

50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

P0hEER _

20.405(a)(1)(1) ,_

Other (Specify 20.405(a)(1)(ll) 50.36(c)(2) ,,__ 50.73(a)(2)(vii) _

LEVEL 0l0l0 50.73(a)(2)(1) 50.73(a)(2)(viii)(A) in Abstract (10) ,_

20.405(a)(1)(lii) ,,_,,,

below and in 50.73(a)(2)(ll) 50.73(a)(2)(vili)(B)

/'/ //////////// ///,///// 20.405(a)(1)(iv) 50.73(a)(2)(lii) 50.73(a)(2)(x) Text)

// // / /// / ,_,_

20.405(a)(1)(v) ,,_,_

f LICENSEE CONTACT FOR THIS LER (12)

TELEPHONE NUISER __

Name AREA CODE 8 l 1 15 31 51 71 -l 61 71 61 1 Don Crow 1. r; xlatory Assurance extension 2960 COWLETE ONE LINE FOR EACH C0fFONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC. REPORTA8LE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE TURER TO NPRDS TURER TO NPRDS N I I I I I I l D I I I I I I I I I I I I I I I l l 1 I l i Expected Month I Day l Year SUPPLEMENTAL REPORT EXPECTED (14)

Submission

~"" lyes (if ves, s_-lete EXPECTED SUBMISSION DATE) TlNO l ll ll l

ABSTRACT (Limit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16)

On Novenber 6,1989 at 1919 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.301795e-4 months <br /> with Unit 1 Defueled, a Primary Containment Isolation System (PCIS) Group 2, 6 and 7 isolations occurred while clearing an out-of-service in accordance with LaSalle Adninistrative Procedure LAP-900 4, " Equipment Out-of-Service Procedure." The PCIS Group 2, 6 and 7 resulted in closure of IVP113A Drywell Cooler 1A Inlet inboard Isolation valve, IVP114A Drywell Cooler IA Outlet inboard Isolation valve, the All other trip of IVP0lPA Primary Containment Chiller pop and IVPOICA Water Chiller unit on low flow.

components were isolated and out-of-service due to scheduled refueling outage work.

The PCIS Group 2, 6 and 7 isolations occurred because the Trip Output and Status switches on the Rosemount reactor level transmitter trip units were not in their required position. When the outage was cleared, junpers were removed that bypassed the contacts for the isolation logic and resulted in an isolation.

l The Trip Status and Trip Output switches were repositioned and the PCIS logic was reset without any further problems.

The Primary Containment Cooling Water system was on at the time to support work in the drywell and was not required for plant operation during this event.

The event is being reported pursuant to the requirements of 10CFR50.73(a)(2)(lv) due to the actuation of an Engineered Safety Feature system.

j 4- .

-' - LifM8 EVENT REPORT (LER) TEXT CONTINUATION Form Rev 2.0 DOCKET NUPSER (2) LER HUfGER (6) Pane (3)- ,

FACILITY NAfE (1)

Year //J Sequential f

/ Revision g/T/ thster l

/7/ thaber

% 11e County Station unit 1 oISloIo10131?l3 sIe - 01216 - 01 0 of2 0F 11 o TEXT: . Energy Industry Identification System (Ells) codes are identified in the text as [XX]

PLANT AND SYSTEM IDENTIFICATION -

J Qeneral Electric - Bolling blater Reactor Energy Industry identification System (Ells) codes are identified in the text as (XX).

l i

4 A. CON 0!T10N PRIOR TO EVENT Event Date: 11/16/89 Event Time: 1919 Hours Unit (s): 1 Defueled Mode (s)Itame: Defueled Power Level (s): _0L, Reactor Mode (s):

8. DESCRIPTION OF EVENT On Itovember 6,1989 at 1919 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.301795e-4 months <br /> with Unit 1 Defueled, a Primary Containment isolation System (PCIS, PC) [JM] Group 2, 6 and 7 isolation occurred while clearing an out-of-service in accordance with LaSalle ,

fdninistrative Procedure LAP-900-4, " Equipment Out-of-Service Procedure." This was done following the-replacement of the Static-0-aing reactor level switches with Rosemount level transmitters and trip ,

units. The PCl$ Group 2, 6 and 7 resulted in closure of IVP113A Drywell Cooler IA Inlet inboard Isolation valve, IVP114A Drywell Cooler 1A outlet inboard Isolation valve, the trip of IVPOIPA Primary Containment Chiller pump and IVP01CA Water Chiller unit on low flow. All other components were Isolated and out-of-service due to scheduled refueling outage work.

Prior to the event Unit I was defueled for a planned refuel outage. The Residual Heat Removal (RNR, RH)

[80] Shutdown Cooling system was isolated and the Primary Containment ventilation loop 1A was in operation to support work activities in the drywell during the outage.

b On October 28, 1989, all outages associated with the Reactor Level Static-0-Ring switch replacement were.

authorized to be placed back in service following the completion of this work.

The Unit 1 Outage Shift Control Room Engineer (SCRE) and an Auxiliary Shift Supervisor (both Ilconsed Senior Reactor Operators) both reviewed the outages and detennined the sequence required to place the new Rosemount reactor vessel level transmitters back in service. The SCRE developed a written guideline to aid in coordinating the effort for clearing the outages and prevent inadvertent Engineered Safety Feature system (ESF) actuations. The guideline for clearing the outages was written to coordinate partially clearing the outages to allow testing of the new Rosemount level transmitters in preparation for reactor vessel flood-up following the vessel decontaminal .on.

The electrical construction testing was congleted for the new reactor level transmitter logic installation, which provides input to both the Reactor Protection System (RPS, RP) [JC] and the Primary Contaltunent Isolation system (PCIS) [JM] per Electrical Construction Test Procedure, ECTP 19-Control Circuits. This test involved inspecting the new devices installed per the modification for physical damage, proper labeling and tenninations.

Form Rev 2.0

- - LIC5K55 EVENT REPORT (LER) TEXT CONTINUATION Pane (3)

DOCKET NumER (2) LER NUSER (6)

FACILITY MAE (1) Sequential Revision Year /

g//

/ Number

/g/// Number L* lle County Station unit 1 015I01010131113 8I9 - 01216 - 01 0 01 3 0F 110 TEXT Energy Industry Identification System (Ells) codes are identified in the text as [XX]

8. DESCRIPT10N OF EVENT (Continued)

On October 28, 1989 the Operating Department partially returned to service the outages associated with 18214403A/8/C/D reactor water level transmitters. This was to provide power the 18214402A/B/C/D and to the master trip unit meters for calibration per LaSalle Instr aent Surveillances LIS 48-101A/8, " Unit i Reactor vessel Low Water Level 3 Scrm Trip Logic Al & A2 and RHR (Shutdown Cooling Mode) Isolation Refuel Calibration" and LIS-MS-107A/S, " Unit 1 Reactor Vessel Low Water Level 1 and Level 21 solation Instrumentation Chanael Calibration." Only a partial calibration was performed to verify the meters associated with the Master Trip Units would respond correctly to a given input signal; no functional check of the logic was completed at this time. Another partial return to service was authorized to allow the IB214402A/8/C/0 and 18214403A/B/C/D level transmitters to be backfilled and locally calibrated using a portable dead weight test ping per LIS-NB-107A/B. Again only a partial calibration This was perfonned to verify the level transmitter would respond correctly to a given input signal.

partial calibration process was done to allow verification of proper instrument trend response when the reactor vessel was refilled. Coupletion of the calibration required removal of all outages.

The Instrument Maintenance personnel coupleted their partial calibratlofts and on Novenber 4,1989, the Operating Department commenced flood-up of the Unit I reactor vessel in accordance with LaSalle Operating Procedure LOP-FC-17, "fliling the Reactor Well and Dryer / Separator Pit from the Suppression Pool through the RHR or Low Pressure Core Spray (LPCS, LP) (BM] System." During the flood-up Technical Staff and Instrument Maintenance Department personnel verified that the newly Installed Rosemount level instrumentation was responding correctly. No problems were found at this time.

Once the Unit I reactor vessel was filled to approximately 12 inches below the vessel flange all the Static-O-Ring level switch replacement outages were authorized to be completely cleared or returned to service in accordance with LAP-900-4. The isolation boundaries associated with the outages were returned to service in the following sequence:

1. The mechanical portion which consisted of the instrisnentation valves.
2. The power supplies to the new Rosemount level transmitters and logic.
3. The retennination of leads for the new instrumentation logic.
4. The removal of jumpers to prevent actual isolation during the replacement of the Static-0-Ring switches.

Prior to the completion of the clearing of the Static-0-Ring level switch replacement outages, the new Rosemount master trip units (a trip unit with a level indicating meter which receives its input directly from the level transmitter) were verified indicating upscale and all trip indicating langs were de-energi zed. Also no PCIS logic status lights were energized prior to clearing the outages. During the removal of the junper on panel IPA 14J, Division 21 solation Logic Auxiliary Relay Cabinet, at terminal points AA-15 to D0-89, a Group 2, 6 and 7 PCIS isolation occurred. Thisjumperbypassedthe Division 2 PCIS Group 2 inboard isolation valve logic. Because the Rosemount master and slave trip units (The slave trip units do not have a level indicating meter and receive their level input from the

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  • '* LirEEEE EVENT REPORf (LER) TEXT CONTINUATION Fons Rev 2.0 DOCKET NUMBER (2) LER NU SER (6) Pace (3)

-FACILITY NAE (1)

Year // Sequential //jj

/ Revision jg//

/ Neber j//

T Naber 0i216 01 0 01 4 0F 11 0 L W 11e County Station unit 1 015101010131713l B I 9 - -

TEXT Energy Industry Identification System (Ells) codes are identified in the text as [XX]

B. DESCRIPTION OF EVENT (Continued) master trip unit.) were not coupletely calibrated, it was not identified that the slave trip units were  !

in t h tripped condition when the juper was removed. (See attached diagram.) Upon further Investigation the master trip units associated with reactor level 3 trips were also found in the trip condition. This prevented resetting the-logic, alarms and resulted in the operation of valves associated with the RPS and PCIS logic.

When the Division 2 PCIS Group 2 isolation logic initiated it resulted in closure of IVP113A Drywell '

Cooler IA Inlet Inboard Isolation valve, IVP114A Orywell Cooler 1A Outlet Inboard Isolation valve, the trip of the IVP01PA Primary Containment Chiller pop and IVPOICA Water Chiller unit on low flow. Prior to resetting the isolation logic all the relay panels associated with the PCIS and Reactor Protection System logic were inspected to determine which relays were energized or de-energized. The following relays were found in their tripped de-energized state:

a. Relays IC71-K6A/B/C/D Reactcr Level 3 RPS and PCIS Group 6 and 7 logic
b. Relays 1821H-Kl8 & C Reactor Level 2 PCIS logic PCIS Group 6 and 7 logic status lights were energized.

These relays were de-energized due to the outage to support the Static-0-Ring switch replacement.

Automatic initiations of Engineered Safety Feature conponents were prevented by junpering across contacts that would normally open when these relays were de-energized. When power was restored to these '

relays while clearing the out-of-services associated with them, the relays should have energized -

providing all trips associated with opening the contacts which supply power to these relays were reset.

Once the relays were energized and all associated contacts that were jupered are closed, the junipers could have been removed because they were no longer necessary.

De-energizing relays 1C71-K6A/B/C/D result in the following actions under nonnal circunstances:

a. A full reactor scran and PCIS Group 6 and 7 isolations and Panel 1Hl3-P601 Control Room alanns B203, "CHAN Al REACTOR AUTO SCRAM" B505, "CHAN A1/B1 RX VESSEL WTR LVL 3 LO" B211, "CHAN A2 REACTOR AUTO SCRAM" B509, "CHAN A2/82 RX VESSEL WTR LVL 3 LO" B303, "CHAN 81 REACTOR AUTO SCRAM" B311. "CHAN 82 REACTOR AUT9 "UAM"
b. Panel IH13-P601 RPS bus A and C F_ran solenoid group lights de-energized.
c. Process Computer alanas D978, "LO RX WTR LEVEL DIV Al" 0985, "LO RX WTR LEVEL DIV A2" 0979, "LO RX WTR LEVEL DIV Bl" 0984, "LO RX WTR LEVEL Div B2"

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I 'N LICi"if EVENT REPORT (LER) TEXT CONTINUAfl0N Fom Rev 2.0

+*

Pee (3)

FAClyTY MAfE (1) DOCKET NumER (2) LER NUMER (6)

Revision

//f Year- f W/

Sequential /,/,/

Naber W/ Naber i

% 11e County Station unit 1 o I 5 I o I o I o 1 31 Tl 3 sIg - oi216 - oIo of 5 0F 11 o TEXT Energy Industry identification System (Ells) codes are identified in the text as [XX] ,

B. DESCRIPTION OF EVENT (Continued)'

d. Control Room back panel PCIS status lights 0531, 33 and 36 energized.

De-energizing relays 1821H-K1B and C result in the following actions under nomal circumstances: I

a. .PCIS Group 2, 3, 4 and 5 isolations Panel IH13-P601 Control Room alarms t l

F504, "CHAN A1/A2 MSiv ISOL TRIP" E504, "CHAN B1/82 MSIV ISOL TRIP" t

- b. Panel 1H13-P603 control Room alams 9511, "CHAN A2/82 RX VESSEL WTR LVL 1 LO-LO-LO" A210. "CHAN A1/B1 RX VESSEL WTR LVL 1 LD-LD-LO" The Operating personnel completed clearing the outages for the Static-0-Ring reactor level switch -

replacement without any further problems. Once the inspection of all the relays associated with the RPS and PCIS logic was coupleted and the outage was coupletely cleared, the Operating personnel attempted to reset the PCIS logic; no change in relay status was observed. The Shift Engineer (SE, Ilconsed Senior Reactor Operator) instructed the Operating personnel to prepare and hang another outage to defeat the PCIS Groups 2, 3, 4 and 5 isolation logic from initiating due to troubleshooting the reactor level 2 logic. The Instriment Maintenance personnel were then requested to troubleshoot reactor level 2 and 3 l

> Rosemount = trip units to detemine why the Operator was unable to reset the logic. After performing a

[

t calibration and inspection of the master and slave trip units it was discovered that the Trip Status switch 51 and Trip output switch 52 located on the printed circuit board for these trip units were in the wrong position. (See attached diagram.) The Trip Status switch 51 was found in the "nomal" position (should have been in the reverse position) and the Trip Output switch S2 was found in the ^

" reverse" position (should have been in the normal position).

The Trip output switch in the " normal" position, controls the contact in the logic for the Master or l'.

l Slave. trip unit which opens to de-energize the RPS or PCIS relay that will initiate a scram or l' isolation. The Trip Status switch in the " reverse" position, contmis the trip indicating LEO located on the master or slave trip unit which should energize to indicate a trip during a trip condition. The Trip Output and Trip Status switches operate independently of each other, therefore the Trip Status light (LED, Light Emitting Diode) may not be lit when the Trip Output relay is in the trip condition.

The setting of the Trip Output and Trip Status switches are all a function of whether the system is energized to initiate or de-energize to initiate. The normal configuration is set to de-energize to initiate.

LICENSEE EVENT rep 0RT'(LER) TEXT CONTINUATION Fom Rev 2.0 LER NUMER (6) Pane (3)

FACILITY NAE (1); DOCKET NUMER (2)

Revislon Year /,/f Sequential W/ Number

//f j/// Nueer 01216 01 0 01 6 0F 11 0 0 1 5 1 0 1 0 1 0 1 31 71 3 819 i dalle County Station unit 1 -

TEXT Energy Industry identification System (Ells) codes are identified in the text as [XX]

8. DESCRIPTION Of EVENT (Continued)

The Trip output and Status switches were repositioned and calibrations conpleted in accordance with LIS-MS-107A/8 and LIS-NS-101A/B without any further probisms.

This event is reportable pursuant to the requirement of 10CFR$0.73(c)(2)(iv) due to an actuation of an ESF system.

C. APPARENT CAUSE OF EVENT The root cause of this event was that the nodification was installed with the Master and Slave trip unit Trip Output or Status switches improperly set. No written instructions were provided in the ,

modification package on how to set the switches. This caused the person in charge of the work to authorize restoring the level instroentation to service in an abnonnal condition. This allowed the level indicating meters on the Master Trip units to indicate upscale with the trip contacts open in the L

tripped condition with no trip indicating Ilght lilisninated on either the Master or Slave trip units.

When the Master and Slave trip units were inspected prior to clearing the outage no trip lasps were jl- 111isnineted. The jumpers were removed across the open contacts while clearing the outage resulting in de-energizing trip logic relays for reactor level 2 and 3 RPS and PCIS logic. This caused PCIS Group 2, 6 and 7 isolations to occur.

A contributing factor was that the Instrisment Maintenance personnel were unable to perfom a complete calibration of the new Rosemount reactor level Instrisnentation and logic until the outage for installing this instrumentation was cleared. Partial calibrations were perfomed on the level indicating meter and transmitters to verify proper response during the reactor vessel floodup. No functional testing of the level instrumentation logic was perfonned at this time, therefore alams and relays were not tested l preventing the Instrument Maintenance personnel from identifying a potential problem prior to returning the Rosemount reactor level instrumentation to service.

An additional contributing factor was the Operating personnel clearing this outage were not required and l

did not utilize the electrical schematics associated with preparing the outage to detemine which alarms l

l should have cleared in the Control Room and which relays should have energized. The outage checklist did not provide special instructions to the Operators clearing the outage to identify potential problems I that could occur when clearing the outage, such as alams not clearing or relays not energizing. This could have allowed them to discover that a trip existed prior to removing the jisuper which bypassed the trip.

Another contributing factor was the Operating personnel did not realize what modification or post maintenance testing was required for the new Rosemount reactor level instrumentation. They were under the impression the testing required for this modification was done and would operate properly when returned to service.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Fons Rev 2.0 1.ER 1RpWER (6) Pane (3)  ;

FACILITY IIAfE (1) DOCKET NUPSER (2)

Year //,/, Sequential ///

W/

Revision thaber W/ thaber

% 11e County Station unit 1 olsIoIoIo 1 31 71 3 e19 - 01216 - oI o 017 0F 11 0 TEXT Energy Industry identification System (Ells) codes are identified in the text as [XK]

D. SAFETY ANALYSl$ OF EVENT (Continued)

The Primary Containment Chill Water (PCCW) system provides chilled water to the Primary Containment cooling units to meet the cooling load requirements in the drywell. The system cooling capacity is based on heat loses from piping and valves, equipment, reactor pressure vessel, and unidentified steam leakages.

The safety significance of this event is minimal since the unit was defueled. At the time of the isolation, most of the equipment affected was already in the isolated condition due to scheduled outage work that was in progress. This type of modification work would not be done in other than outage conditions. The problem with the trip units would have been corrected prior to declaring these instroents operable.

E. CORRECTIVE ACTIONS I The Operating personnel on shift prepared another outage in accordance with LAP-900-4 to bypass RPS and PCIS reactor level 2 and 3 trips. The outage was conpleted and logic reset.

1 Once the PCIS logic was reset the IVP113A Drywell Cooler IA Inlet inboard Isolation and IVPil4A Drywell Cooler lA outlet inboard Isolation valves were reopened. The IVPOIPA Primary Containment Chiller Pung and the IVPOICA Water Chiller unit was restarted.

l The Instroent Maintenance Department was requested to troubleshoot and calibrate the Rosemount reactor l

level instraentation in accordance with LIS-MS-107A/B and LIS-NB-101A/B. During the perfonnance of the l

I calibration the Instroent Maintenance personnel determined the problem was the 1821-N703A/8/C/D Master and 1821-N7048/C Slave trip unit Trip output and Status switches being in the wrong position. The Trip Output and Status switches for each of the trip units were repositioned and the calibration was conpleted without any further event, A review will be done to detennine if additional training is needed on vessel instrumentation and r

l available indications which can aid Operating personnel in the Control Room in detennining the actual status of RPS or PCIS logic. Action item Record (AIR) 373-200-89-10801 will track the progress of this review.

A revision of the LaSalle Adninistrative Procedure 1AP-900-4, " Equipment Out-of-Service Procedure" will be done to include the following revisions:

(1) Add a caution to consult with the Technical Staff or the working department for assistance in preparation, installation and removal of outages when technical expertise is required to prevent an Engineered Safety Featured system actuation. This revision will assist in developing special instructions when clearing an outage.

(2) Provide more guidance to the person in charge of the work to insure the work is completed such that equipment damage or an Engineered Safety Featured system actuation would not result when the outage is cleared or equipment is restored to service, if any special instructions are required the working department must provide them to the Operating Department.

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LIC5aKEE EVENT REPORT (LER) TEXT CONTINUATION Fors Rev 2.0

. i DOCKET NumER (2) LER NUMER (6) Paee (3)

. FACILITY NAM (1)

Year /y/ Sequential //g Revision f

/// Weber /// Neber

% 11e County Station unit 1 0 1 5 1 0 1 0 1 0 1 31 71 3 e19 - 01216 - 01 0 01 8 0F tl 0

. TEXT' Energy Industry Identification System (Ells) codes are identified in the text as (XX]

E. CORRECTIVE ACTIONS (Continued)

(3) . The person in charge of the work wl11 be required to insW Wt testing of logic has been completed, when possible, prior to the authorization of fu 4 (aturning the equipment to service. Consideration should be given to partial clearing or temporary lifting the outage to

  • acconpilsh this. Any untested logic must be reported to the Operating Department.

I (4) Concerning outage clearances, including temporary lifts and partial clearances a statement will be added to consider returning to so vice one channel at a time. This is to accospilsh I testing or allow the system to be placed back in service to troubleshoot without causing an Engineered Safety Featured system actuation.

AIR neber 373-200-89-10802 will track the progress of this review.

' A General Information Notification (GIN) will be developed to review this event with Operating,

' Maintenance, Substation Construction, Operational Analysis Department, Engineering and Construction

' Department and Technical Staff personnel. AIR neber 373-200-89-10003 will track the progress of this review.

F.- PREVIOUS EVENTS LER Nunber Title 373/64-015-00 Inadvertent Group 2 and 4 Containment isolation 373/96-011-00 Personnel Error Primary Containment Isolation 373/06-037-00 Group 2 and Group 4 Contalrunent Isolation Due to inadequate Out-of-Service 374/87-006-00 Unplanned Engineered Safety Feature Actuation During Modification Testing caused by Personnel Error G. COMPONENT FAILURE DATA None.

4

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Year /H Sequential /n Revision ,

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tah lte county Statlan unit 1 015101010131713 819 - 01216 - 01 0 01 9 0F 11 0 <

TEXT Energ ladustry lesatification System (Ells) codes are identified in the text as [XX]-

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IJCNESEE EVENT REPORT (12R) TEXT CONTINUATION Form Rev 2.0 12R NUMBER (6) Page ll3)

FACI 1JTY NAME (1) DOCKET NUMBER (2)

Year 7 Sequential W/Revisl0n

' / Number 7/, Number taSalle County Sta'llon Unit 1 0l5l0l0l0l3l7l3 8l9 -

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0l0 1l0 0F l' l 0 TEXT Energy Industry identification System (EDS) oodes are identified in the text as (XX}

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