ML20043F216

LER 89-024-01:on 891214,main Turbine Unexpectedly Tripped Following Isolation of Reactor Water Level Switch (Lits). Caused by Channel B Lits Having Been Previously Replaced W/ Switch That Operated Reverse function.W/900601 Ltr

ML20043F216
Person / Time
Site:	Quad Cities
Issue date:	06/01/1990
From:	Bax R, Michael Brown COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-024, LER-89-24, RLB-90-138, NUDOCS 9006140260
Download: ML20043F216 (11)

Text

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t . Commonwealth Edison

. Quad Cities Nuclear Power Station 22710 206 Avenue North oorcova, Illinois 61242 -

Telephone 309/654 2241 RLB-90-138 June 1, 1990 U. S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555

Reference:

Quad Cities Nuclear Power Station Docket Number 50-254, DPR-29, Unit One Enclosed is Licensee Event Report (LER)89-024 Revision 01, for Quad Cities Nuclear Power Station.

This report is being submitted as a voluntary report.

Respectfully,

'COMMONNEALTH EDISON COMPANY QU CITIES NUCLEAR POWER STATION R. L. B '

Station Manager RLB/MJB/jlg Enclosure cc: R. Stols T. Taylor INPO Records Center NRC Region III i

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4 LICENSEE EVENT REPORT (t[R)

Facility Nahie (1) Docket Number (2) _D oe f3) cuad Citiet Unit One of El 01 01 01 21 El 4 I of 1 0 Title (4)

U1 Turbine Trio Due to Hioh Rx < Water Level 11cnal Caused by und.glected Trinced Level Switch

, Event Date fE) LER Number (6) Recort Date (7) Other Facilitiet Involved (B)

Month Day Year Year //p/ Sequential f//jj/ Revision Month Day Year Facility Names l Docket Numberft)

/// Number /// Number el El 01 01 of I l

~~ ~"

_11 2 11 4 al 9 81 9 0 12 l4 0l 1 0 l6 011 91 0 Di El 01 0! 01 i i THIS REPORT !$ $UBMITTED PUR$UANT TO THE REQUIREMENTS OF 10CFR f Check rie or more of the f ollowino) (11) 4 20.402(b) _ 20.405(c) _ 50.73(a)(2)(iv) _ 73.71(b)

POWER , 20.40$(a)(1)(1) 50.36(c)(1) __._ 50.73(a)(2)(v) _ 73.71(c)

LEVEL 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) 1 Other (specify lE (101 0 3 20.405(a)(1)(iii) _ 50.73(a)(2)(1) 50.73(a)(2)(viii)(A) in Abstract

/// /& // /////// // / 20.405(a)(1)(iv) 50.73(a)(2)(11) 50.73(a)(2)(viii)(B) below and in

/jjj//jj/jgj/jjj/j/jfjjjj%jj/jjj/j SHHHHHHHHHS %% ** **M*n'Hn '*(*H H"') '*(*H'Hn '**U i LICENSEE CONTACT FOR THIS LER f12) i Name TELEPHONE NUMBER AREA CODE f

Mike Brown. Peculatory Assurance. Ext. 3102 3 10 l9 61 El 41 -l 21 21 4l j CDMPLETE ONE LINE FOR [ ACH COMPO FAILURE DESCRIBED IN THIS REPORT (13)

CAU$E SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE TURER 70 NPRDS TURER TO NPRDS i

__ l l l l 1 1 I I I I I I l l I I I I I I I I i 1 I I I I SUPPLEMENTAL REPORT EXPECTED fi41 Expected Month ! Day I Year Submission lYgt fff vet. comolete EXPECTED $UBMISs!Qll.DATEl X f NO l l l ABSTRACT (Limit to 1400 spaces. i.e. approximately fif teen single-space typewritten lines) (16)  !

1 ABSTRACT:  !

On December 14, 1989 at approximately 0315 hours0.00365 days <br />0.0875 hours <br />5.208333e-4 weeks <br />1.198575e-4 months <br />, Unit one was operating in the RUN uode at 35 percent of rated core thermal power. The main turbine unexpectedly tripped following the isolation of Reactor Water Level Switch (LITS) 1-263-59A.

The cause of the trip was a result of the B channel LITS having been previously replaced with a switch that operated the reverse of what was required. Thus, when the A channel LITS was isolated, the turbine trip logic was completed. The alarm that should have been annunciated due to the condition of B LITS, was found to have been inactivated as a result of its signal leads being inadvertently left determinated during a modification installation.

Corrective actions included replacing the B LITS switch, reterminating the alarm leads, checking the other unit, testing, and a procedure revision. Further corrective actions are in process. r 2

This report is being submitted as a voluntary report.

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e LICENEff [ VENT REPORT fLER1 TEXT CONTINUATION _

form Rev 2.0 FACILITY N'AMC (1) DOCKET NUMBER (2) LER NUMBER f61 Pane f31 8 a Year 5eauential / Revision

• ///p/

// Number /p/j/p

// Number _

Ouad Citiet Unit One 0 lEl0l0 l 0 1 21'11 4 819 - 01214 - 011 012 0F 110 itxT Energy Industry Identification system (tIIs) codes are identified in the text as (xx)

PLANT AND SYSTEM IDENTIFICATION:

General Electri - Boiling Water Reactor - 2511 MWt rated core thermal power.

EVENT IDENTIFICATION: U1 Trubine Trip due to High Reactor Water Level Signal Caused by undetected Tripped Level Switch.

A, CONDITIONS PRIOR TO EVENT:

Unit: One Event Date: December 14, 1989 Event Time: 0315 Reactor Mode: 4 Mode Name: RUN Power Level: 35%

L This report was initiated by Deviation Report 0-4-1-89-121 l

l .RUN Mode (4) - In this position the reactor system pressure is at or above 825 l psig, and the reactor protection system is energized, with APRM protection and RBM

, interlocks in 4ervice (excluding the 15% high flux scram).

l l On December 12, 1989, at 2120 hours0.0245 days <br />0.589 hours <br />0.00351 weeks <br />8.0666e-4 months <br />, a Radiation Protection Technician (RPT) cbserved water dripping from the Unit One Level Switch (LITS) [LTS]'l-263-59A.

This switch provides one-half of the Turbine [TRB)[TA) Trip and Reactor Feed Pump (RFP) [P) [SJ) High Reactor Water Level trip logic [JC). Hourly monitoring of the-leak was initiated while the work package was prepared for accomp~lishing the repair.

l 1

On December'13, 1989,'at 2320 hours0.0269 days <br />0.644 hours <br />0.00384 weeks <br />8.8276e-4 months <br />, a load reduction wts initiated at approximately 150 MWe/hr to within the bypass valve capability to prevent a reactor scram [JC) in the event of a turbine trip during the repair. The load reduction was to 280 MWe.

The reactor feed pumps high water level trip was taken out of service on December 14, 1989, at 0245 hours0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br /> to prevent an inadvertent trip during the repair.

B. DESCRIPTION OF EVENT:

The' turbine tripped at 0316 hours0.00366 days <br />0.0878 hours <br />5.224868e-4 weeks <br />1.20238e-4 months <br />, on December 14, 1989, following the isolation of the high side of the level switch and opening of the equalizing valve [V) by Instrument Maintenance (IM) personnel.

The trip was unexpected since only one-half of the logic was thought to be made up by this maintenance activity.

The Reactor Operators entered station procedure QOA 5600-4, Loss of Turbine L Generator. They also referenced 00A 3500-1, Loss of Feedwater Heaters, since the l, feedwater heaters (HX) had tripped as a natural result of th? turbine trip, and l -noted that this procedure required a reactor scram if the feedwater temperature decreased by more than 140 degrees Fahrenheit (*F). As the temperature decrease approached 140 'F, the Shift Engineer (SE) made the decision not to scram the reactor based on the knowledge that this requirement was intended for a loss of feedwater heaters from a high power level. The Production Superintendent and the Lead Nuclear Engineer were consulted and concurrence was received not to scram.

The concern with this condition is the potential for violation of the Minimum

( Critical Power Ratio (MCPR) safety limit. Since reactor power had been reduced to 280 MWe prior to the transient, MCPR was not a concern.

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i= . ._

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Fev 2.0 FACILITY NAME (1) DOCKET NUMBtR (3) LER NUMBER f6) Pace (3) _

' seguential /j/j/ Revision j/jj Year //

/// Number j//

/ Number

.jggd Citiet Unit One 0 l$10l0l0l 21 El 4 8l9 0l 214 - 0 l1 013 0F 110 TEXT Energy Industry Identification system (E!!s) codes are taentified in the text as (XX)

At 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />, the Production Superintendent notified Corporate CECO of the event.

A mutual decision was made not to shutdown the unit. The decision was made to conduct a corporate review of this event.

A conference call between the corporate office and the station conducted later in the day raised questions with respect to the design casts, since the feedwater temperature decrease of 145'F to 150*F had exceeded the 145'F assumed in the reload analysis. It was decided to shutdown the unit and make an Emergency Notification System (ENS) phone notification since it could not be conclusively determined that the unit was within the design basis. General Electric was contacted and indicated that the event was bounded by the design analysis. The shutdown was terminated based on General Electric's evaluation and the ENS notification was rescinded.

The station's initial investigation of the cause of the turbine trip determined I that work request 078956, performed during the previous refueling outage, had incorrectly open rather,thanreplaced normal switch closed #4contacts.

of LITS 1-263-59B with in This resulted a switch that being the switch had normally in the tripped condition with the reactor water level normal. When LITS 1-263-59A was isolated for maintenance, this completed the turbine trip logic.

The station also investigated the cause for the RFP/ Turbine Trip High Level alarm

[ ALM] not being up. This alarm should have warned the operator that one-half of the turbine trip logic was activated. Actuation of either the LITS 1-263-59A or B cill activate the alarm. The investigation determined that the leads [JX) to this annunciator (ANN) had been inadvertently left determinated during installation of the Detailed Control Room Design Review (DCRDR) modification for Panel (PL] 901-8 during the previous refueling outage.

A restart plan was formulated and an on-site review conducted to authorize restart of the unit. These actions corrected the identified equipment problems, performed tests to ensure other similar problems did not exist, and revised QOA 3500-1, to include additional guidance for the operators. The General Electric evaluation of the loss of feedwater heaters analysis was included as part of the on site review.

Corporate CECO and NRC concurrence for restart was received. The Unit was resynchronized to the grid at 0619 hours0.00716 days <br />0.172 hours <br />0.00102 weeks <br />2.355295e-4 months <br /> on December 15, 1989.

This event was considered to be a Potentially Significant Event (PSE) and a

preliminary PSE report 89-11, was issued on December 18, 1989.

, C. APPARENT CAUSE OF EVENT:

This Licensee Event Report (LER) is being submittru as a voluntary report. This report also serves as the final review for the PSE.

The main cause of this event is attributed to an improper switch being installed in I

LITS 598. The source of this problem was traced back to a modification installed l on May 4, 1972. This modification changed switch #4 of LITS 1-263-59A and B from normally open to normally closed. The vendor manual was not revised to reflect this modification and this information was not included in the maintenance

{ history. There was not a requirement to update vendor manuals and the requirements for maintenance history were considerably different in 1972.

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e LICENitt IV[Tj_g[ PORT f LtR) TEXT CDNTigl4TIDN Form Fev 2.0 FACILITY NAML (1) DOCKET NUMBER (2) LtR NUMBtR (6) Pane (3)

Year //p/ sequential

/pp//l Revision

/// Nu ger p/// Number

_0uad Cities Unit One 0 l5 l0 10 l 0 1 21 El 4 8 10 - 01214 - D 11 014 0F 110 TEXT Energy Industry Identification system (E!!s) codes are identified in the text as [XX) ,

The incorrect switch was installed by work request Q76428 on November 11, 1989.

The vendor manual was utilized to determine the correct part number for the .

switch. Although the vendor manual was controlled under the VETIP program, this  !

error in the manual was not detected as part of the review process.

Several barriers were available to prevent installation of the incorrect switch.

The IM who installed the incorrect switch noted that the part number was different from the part number of the switch removed and notified his supervisor. The supervisor utilized the vendor manual to verify the correct part number, and I therefore did not identify the problem. The Post Maintenance test (PMT) was inadequate to verify operability of the trip circuit. The EHC line-up is a separate test, not required by Technical Specifications, to determine turbine trip operability prior to start-up following a refuel outage. The sequence in which the Electro-hydraulic Control (EHC) [TC) Lineup Test was performed did not allow I detection of the oroblem. A Parts Technical Evaluation also failed to identify the l problem since the part was considered like-for-like and a suitability of  !

application was not required. The Parts Technical Evaluation utilized a print for a switch which the vendor manual showed as requiring the same part number and not the actual print for the switch in question. A verification of relay status for the B channel prior to inserting a trip in the A channel would have prevented the turbine trip, but it was not performed and not required by procedure.

Contributing to this event was the inoperable RFP/ Turbine Trip High Level annunciator. The leads for the RFP/ Turbine Trip high level Annunciator, F-11 on panel 901-6 were lifted on October 17, 1989 in accordance with work package 076430 for modification M-4-1-87-518. This package is one of eight work packages that implement the wiring changes for DCRDR.

These leads were lifted in the above work package for panel 901-8 as a result of I the process utilized by the contractor work analyst to construct the work package. l The work package was constructed to accomplish all wiring changes in panel 901-8 i necessary to place the panel in the final as built configuration called for by the i design. When the design called for termination of new wires on terminals that were already utilized, the practice was to determinate existing leads with the knowledge that they would be reterminated at a later point in the modification package. This design approach failed to consider that the modification would not be entirely completed during the refuel outage. The work analyst identified the potential of inoperable annunciators on panels not to be completed during the refuel outage and communicated his concerns to Engineering and Construction Services (ENC)-Construction via a memo. He was verbally instructed by ENC-Construction to l delete.the determination of these and other similar cables from his work package.

In accomplishing this task, he neglected to delete the determination of the two leads in question from his work package.

As a result of the above error, a detailed review of the M-4-1-87-051B work packages was initiated to determine if other problems existed due to the above l

approach. This review identified five additional annunciator operability concerns. Three of these concerns had been corrected as a result of additional '

t* portions of the modification being completed. The fourth concern was corrected via a work request on December 18, 1989. The fifth was corrected during post-modification testing.

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- 'I

. Form Rev 2.0 ,

LitENSEE EVENT REPORT f LEP) TEXT CONTINUATION LER NUMBER f61 Pane f3)

.kACILITYhAME*(1) 00tKET NUMBER (2)

Year seguential / Revisitin I- p/p//

f Number

,////,/ Number _

///

0 l2 l4 011 ole OF 1 10___

Guad Citiet UniLQne 0lE l.0 l 0 l 0 121514 819 - -

h TEXT (nergy industry identification system (Ells) codes are identified in the text as [XX)

During the review of this event, it was determined that the portion of the modification that renumbers all of the terminal points had been completed, Thus, the as although,' rewiring had not been completed for all of the panels.

built drawings for these panels did not reflect the as labeled terminal numbers.

Several barriers were available to prevent this error. The initial review of the work package by the contractor, ENC-Construction, and the. station should have been sufficiently detailed to detect this problem, The Operational Analysis Department (OAD) test should have detected this error, but it did not, due to the use of schematic diagrams for development of the scope of the test. The schematic diagrams provided no indication that the leads in question had been disturbed, this information was available in the work package. The modification test and operability test were performed in conjunction with the OAD test,'and therefore also did not identify the error. The user's walkdown performed by the station was a general inspection for material condition and housekeeping and did not include a 100 percent wiring verification. The station utilized the 100 percentThis wiring verification performed by the installer to satisfy this requirement.

verification utilized the work package and not the design drawings, therefore, could not have detected the error since the work package specified that the leads were to be determinated.

D. SAFETY ANALYSIS OF EVENTj, To mitigate an Abnormal Operating Occurrence (A00) involving a a postulated reactor coolant level increase. a set of two switches located at the high water level (HWL) setpoint, +48 inches, are designed to provide signals which trip the feedwater pumps and initiate a turbine trip upon indication of a HWL condition. The failure of one switch during a postulated A00 such as feedwater Controller Failure -

Maximum Demand (FWCF) could lead to a situation in which the reactor water level would continue to increase past the high water level setpoint with no trip of

-either the feedwater pumps or the turbine stop valves.

A transient safety evaluation was performed and demonstrated that in the event of a postulated reactor coolant increa:0 ADO such as the FHCF Maximum Demand event occurring at rated power / flow condition, the failure of the HWL trip function to shut down the feedwater pumps and to initiate a turbine trip and subsequent anticipatory scram signal would not result in any adverse impact Theto the plant's safety evaluation operating MCPR' limits or the fuel thermal-mechanical limits.

has demonstrated that such an event remains bounded by the limiting reload licensing basis event. Therefore, the safety consequences of this event were minimal.

For further safety analysis information, reference G.E. Nuclear Energy Letter EB0-90-079, dated February 13, 1990, to Dr. D. F. Naughton. This information was reviewed and accepted by Nuclear Fuel Services and the Station.

L E. CORRECTIVE ACTIONS:

PRIOR h Corrective actions for this event are separated into four major categories:

TO TURBINE RESTART, LITS 59B LEVEL SWITCH, INOPERABLE ANNUNCIATOR, AND OPERATIONS / TEMPORARY PROCEDURES.

A composite listing of the corrective actions completed and to be completed are listed.

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I ,

i L R ut1EE EVENT REPDFT_,jLER) TEXT EDNTINUATION Form Rev 2.0 FACILITY NAME (1) DOCKET CUMBER (2) LER NUMBER (6) Pane (3) 6 Year Sequential Revision Number Number

-Quad Citiet Unit One 0l5l0 1 0 1 0 l 21 El 4 8I9 - 01214 - 0l1 016 0F 110 TEXT Energy Industry Identification system (E!!s) codes are tdentified in the text as [xX1 The actions denoted with an asterisk (*) are corrective actions discussed by Commonwealth Edison Company and the Nuclear Regulatory Commission during a meeting held on January 11, 1990.

PRIOR TO TURBINE RESTART Corrective actions I through 6 apply to this area. The following actions have been completed:

1) The Unit Two turbine trip logic was checked and it was verified that the same problem did not exist on Unit Two.  !

2)* Repairs to LITS 1-263-59A were completed and it was successfully tested. l 3)* Switch #4 of LITS 1-263-59B was replaced with a switch with normally closed contacts and successfully tested.

4)* The leads for the F-11 alarm were terminated and tested satisfactorily for '

both switches, 59A and 59B.

5)* QOA 3500-1 was revised to provide additional guidance on the intent of the procedure. Each operating crew was trained on this procedure immediately I following the assumption of shift duties.  ;

6)* An operability test of control room annunciators important to plant operations  !

was completed on panels 901-4, 5, and 6. The annunciators on panels 901-3, 7, and 8 were tested as part of the modification.

LITS 59,8 LEVEL SWITCH Corrective actions 7 through 17 apply to this area. The following corrective j actions have been completed. '

7)* Vendor Manual C00038 was revised to indicate the correct model switch.

i 8)* It is required that the design print in conjunction with the vendor manual and I other applicable information be utilized for Technical Evaluations.

l 9)* A sample review of IM, EM. and MM work packages performed during the outage for which a vendor manual was referenced for parts procurement or post i maintenance testing was completed with no problems noted. j 1

10) IM personnel were counseled on the importance of attention to detail and i follow-up to ensure any discrepancies are adequately resolved.

11)* The EHC Lineup Test procedure was formalized as the QIP 5610 block of procedures to include sufficient detail to ensure desired actions are j.

accomplished.

12)* A. sample of IM and EH work packages performed during the outage was reviewed  ;

for adequacy of post maintenance test, no problems were found.

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'a

[ LICENSEE EVENT REPORT f LfR1 TEXT CONTINUATION Form Rev 2.0 F Ac!LITY 'NAME (1) DOCKET NUMBER (2) LER HUMBER f6) Pace f3) 6 . Year /// sequential /jj j// Revistor.

//,/ Number /// Number

. Quad Cities Unit One 0 l51010 10 l 21 51 4 8l9 - 012 l4 - 011 017 Or 110 TEXT Energy Industry Identification system (E!!s) codes are identified in the text as [XX)

The following corrective actions are to be completed:

13)* Procedurally require replacement parts to be compared to parts removed.

Like-for-like should be verified and documented when used and differences should be reconciled prior to continuing work or considering equipment operable (NTS 2542008912113). i 14)* Require Maintenance and Operations personnel during a non-routine surveillance activity to verify relay state prior to inserting trip signals whenever accessible (NTS 2542008912114).

15)* Review the adequacy of guidance provided for the content of post maintenance test and upgrade as necessary to ensure operability is demonstrated (NTS 2542008912115).

16)* Review the adequacy of tne vendor manual review program with respect to schedule and scope (NTS 2542008912116).

17)* Develop and implement a guideline for utilization of vendor manual information until vendor manual upgrade program is complete (NTS 2542008912117).

INOPERABLE ANNUNCIATOR Corrective actions 18 through 40 apply to this area. The following corrective actions have been completed:

18)* A review of all completed work packages as part of this modification was completed to ensure that additional problems did not exist. Five additional problems were identified. Three of these problems were corrected during the normal course of the modification as the work proceeded to the panel affected. The fourth error was corrected with a work request. The fifth was corrected during post-modification testing.

19) The use of a formal information transmittal system between ENC / Contractor /AE/ Technical Staff for non-safety related modifications is now required.

20) A cross reference matrix of Sargent & Lundy and Impell terminal numbers for all terminal points affected by this modification was developed.

21) All wires that were made spares during this modification were reviewed to verify that they should remain spares. No problems were noted.

22) All FCRs associated with the annunciator modification were reviewed to ensure that they did not create problems. No hardware problems noted.

23)* All modifications completed which utilized an OAD test as the modification test to verify that operability was adequately demonstrated were reviewed with no problems identified.

24)* Partial work packages are now required to contain flags to highlight ties with other work packages.

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LICENSEE EVENT REPORT fLER) TEXT CONTINUATION F orn' Rev 2.,L f At!LITY NAME (1) DOCKET NUMBER (2) LER NUMBER f6) Pane f3) 0 Year /// sequential /p,/p/ Revision p/p,

// Number /// Number _

Ouad Cities Unit One 0 !5l010 1 0 l 21 El 4 8 i 9 - 0 l2l4 - 0 l1 Ola 0F I!L TEXT Energy Industry Identification system (E!!S) codes are identified in the text as (XXI 25)* The use of Nuclear Station Work Package (NSWP) form E-02, Exhibit H.

"DETERM/ SPARE Checklist for Abandoned Cable / Conductors" or equivalent is required for all modifications involving wiring. The checklist is included in the post modification review for operability.

26)* A 100 percent sample of partial and incomplete modifications closed during the outage was reviewed for adequacy of users walkdown, modification test, and configuration control. No problems were noted.

27)* A tailgate review was conducted with contractor's supervision, engineers, work analysts, and Q.C. of the facts, conclusions, and corrective actions surrounding the annunciator event.

28)* The OAD test procedure is now prepa 9d utilizing the work package and design drawing to ensure a comprehensive tes! is performed.

29)* The requirement to use drawings in addition to work instructions to perform walk-downs was reemphasized.

30)* The expectation for the level of involvement of ENC / Project Management with respect to partial modifications was clarified.

31)* Ensure that an independent review is completed prior to issuance of the modification work package to the field.

32)* Procedures have been revised to require that the Tech Staff test will be independent of the OAD test and is prepared utilizing the prints and work packages.

The following corrective actions are to be completed:

33) Ensure that the lessons learned from this event are communicated to other stations and understood (NTS 2542008912133).

34) Perform a review of the modification program and the Architect Engineer (AE)

Guidebook to determine if any changes are necessary (NTS 2542008912134).

35)* Revise the scope of guidelines for ENC / Construction review of DCRDR annunciator mod work package instructions to include interaction with other work packages (His 2542008912135).

36) Technical Staff will be supplemented with three additional contractor personnel until such time as the staffing lev'el and experience are deemed adequate (NTS 2542008912136).

37)* Review all partial and incomplete modification packages to incorporate specific guidelines from the designer as appropriate (NTS 2542008912137).

38)* Clarify cxpectation for the responsibility of ENC / Construction in their J support to the station to achieve error-free maintenance and modification C installation work (NTS 2542008912138).

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f t

• LICEhiEE EVENT REPORT (LFR) TEXT CONTINUATION Form Rev 2.0 L FA(ILITY N'AME (1) DOCKET NUH6ER (2) LER NUMBER f6) Pace f3) l ' ' Year // Sequential / Revision

/pj,/ Number ,/j/j/

// Number

( //

Quad Cities Unit one 015 1010 1 0 1 21 El 4 B l9 - D 1214 - 011 IOlh DF 110 TEXT. Energy Industry Identification System (t!!s) codes are identified in the text as (XX) 39)* Procedures have been revised to ensure that an effective wire by wire users walkdown is completed prior to post modification test (NTS 2542008912139).

40)* The Mod Task Force will review this event and corrective actions for possible inclusion in the modification program (NTS 2542008912140).

OPERATIONS / TEMPORARY PROCEDURES Corrective Actions 41 through 49 apply to this area. The following corrective actions have been completed:

41) The procedure.for temporary procedure changes was revised to bring it into compliance with the Technical Specifications.

42) A review was performed of all active temporary procedure changes to ensure that the requirements of the Technical Specification are adequately addressed.  !

43)* QOA 5600-4, " Turbine Trip", was revised to acknowledge the loss of feedwater heating during the turbine trip and provide guidance to operators concerning expected indications and required actions.

44)* LER 254/89-023 was issued and on-site reviewed and approved.to address the Technical Specification violation for temporary procedure changes.

45)* The event was discussed with operations personnel to ensure that they have not received the wrong message. The company and station policy of conservative operations was reinforced.

46) Additional training on procedure adherence was completed and expectations were reemphasized. 1 The following corrective actions are to be completed:

47)* Expand on-going training on the use of procedures in order to reinforce expectations-(NTS 2542008912147),

48) Ensure that the new simulator adequately models this transient (NTS 2542008912148).

49) Develop an enhanced program for incorporation of Technical Specification '

changes into station procedures (NTS 2542008912149).

50) Upgrade the program for training on procedure revisions (NTS 2542008912150).

F. PREVIOUS EVENTS:

No similar events have occurred on the 1-263-59A and B level switches previously, however deviation investigation report (DIR) 04-01-89-70 was written for an error associated with modification 04-1-87-0518, which caused the Steam Jet Air Ejector 2~ (SJAE) suction valves to close on an operating unit.

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LICENitt EVENT REPORT fLtP) TEXT CONTINUAT10N Foff . ' L.

FAc!LITY NAME (1) DOCKET NUMBER (2) LER NUMBER f6) P. 3 ,, _

+

Year /

j/pj/ sequential /jj/p

/ Revision

/// Number /// Number Dund cities Unit One 0 l1l01010 1 2! El 4 al9 - 0l2l4 - 0 11 Ill 0F 110

' TEXT Energy Industry Identification system (EII;) codes are identified in the text as (XX)

Based on the corrective actions completed and in process, no further action is-deemed necessary.

'G. COMPONENT FAILURE DATA:

There are no component failure associated with this event.

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