IR 05000298/1990026

Insp Rept 50-298/90-26 on 900616-0715.One Unresolved Item Identified.Major Areas Inspected:Operational Safety Verification,Monthly Maint & Surveillance Observations,Esf Walkdown & Onsite Followup of Written Repts

ML20058P420
Person / Time
Site:	Cooper
Issue date:	08/08/1990
From:	Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20058P418	List: IR 05000298/1990026 ML20058P415
References
50-298-90-26, NUDOCS 9008170034
Download: ML20058P420 (14)

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APPENDIX U.S. NUCLEAR REGULATORY C0FNISSION REGION IV- i

-l NRC Inspect 1onl Report: 50-298/90-26 Operating License:~ DPR-46 Docket: 50-298; Licensee: Nebraska Public Power Dist;ict (NPPD)

P.O. Box 499!

LColumbus, Nebraska 68602-0499 Facility Name. Cooper Nuclear Station (CNS),

Inspection At: CNS, Nemaha County,-Nebraska-Inspection Conducted: June 16 through July 15, 1990 Inspectors: G. A. Pick, ~ Resident Inspector, . Project Section C Division of Reactor Projects-R. E. Farrell, Senior Resident Ir.-Dector, Fort:St. Vrain  ;

Division of.: Reactor Projects R. V. Azua, Project Engineer, Project Section C i

- Division' of Reactor Projects-. -l

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W., R. Bennett, Senior Resident -Inspector, Project section C I Division of Reactor Projects

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Approved: (IVhD $)[$10Lf 0 GiL. Constable,Clilef,ProjptSectionC Date '

Division of Reactor Projects

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Inspection Summary Inspection Conducted June 16 through July 15, 1990 (Report 50-298/90-26)

Areas Inspected:- Routine, unannounced inspection of operational safety verification. monthly surveillance observation, monthly maintenance observation,

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onsite followup of written reports, engineered safety features (ESF) walkdown, and in-office. review of event reports.- Within these areas, the inspection. . ,

consisted of selective examinations of procedures and representative records, '

interviews with personnei, and observations by the-inspector Results: -Within the ar6as inspected, an unresolved ite:a was identified i concerning.cn-the-job training of' instrument and control (I&C) technician l This unresolved item is discussed in paragraph d.g of the repor _

I 9008170034 900000 Y PDR ADOCK 05000298F Q PNtjgjj

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i The-licensee plans to complete a' review of fire coors located in the plant by -

.- leptember 15, 1990, to ensure that a proper analysis exists for each door configuratio ~

Thelicenseehasapprovedanupdatedsafetyanalysisreport(USAR)' change increasing the design basis service water inlet temperature to 90'F. The engineering department-has analyzed data- to allow opei stions if temperatures exceed 90 '

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3 DETAILS- Persons Contacted Principal Licensee Employees

• J 1 M._ Meacham, Division Manager of: Nuclear Operations-

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• S. M.- Peterson,- Senior Manager, Technical Support

• H. T.- Hitch, Plant Services, Manager

• J. R. Flaherty, Engineering Manager '

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• R. Brungart 0perations Manager:

• J. V. Sayer, Raoiological Manager

• M. E. Unruh, Acting Maintenance Man &ger

• L. E. Bray. Regulatory Compliance-Specialist The inspectors also interviewed other! licensee employees during the inspection perio ,

• Denotes those present during tne exit. interview on July 17,;199 ' Plant Status The plant operated'at essentially 100 percent power.from June 16 through t

July 6, 1990. The licensee decreased. power to 154 MWe on July 7 t balance Bearing No.- 5 on, Low. Pressure Turbine No. 2. The unit returned to full power on _ July 9.and remained at. essentially 1100 percent power until the end of. the inspection: perio . Operational Safety Verification (71707) - The.' inspectors observed operationalLactivities throughout the ..

inspection period. Contml room activities-were observed to be well'

controlled. Proper control room staffing was maintained;and :

professional conduct'was continuously observed. Disc'ussions with operators determined-that they were cognizant:of. plant status and understood the importance of, and reason for; each lit" annunciato Operators implemented the required heat balance calculations' and thermal limit surveillances. :The . inspectors.. observed selected shift turnover meetings and noted -that information concerning' plant status and planned evolutions was communicated to the oncoming operator The inspectors verified daily, by visual inspection of emergency core

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cooling system valve ~ position indicators, that the' systems were maintained in a standby conditio The inspectors verified that' selected activities'of the licensee's-radiological protection program were implemented in conformance with

facility policies, procedures, and regulatoryj requirements'.: Radiation

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and/or contaminated areas were-properly posted and controlled. ' Health physics (HP)- personnel were prompt in re)osting radiation areas !

affected by the shutdown. Radiation wor ( permits contained

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appropriate information' to ensure that work could be performed in a: "

safe and controlled manner. - HP personnel were. observed to be touring 3 work areas, ensuring proper implementanon of- as low as reasonabl achievable and radiological' control requirements. . Radiation monitors were properly utilized to. check for contaminatio /

The inspectors observed security personnel perform their duties c'f - -!

vehicle, personnel, and package search. Vehicles were properly: .

authorized and' controlled or escorted within' the protected area.. The inspectors conducted-site tours.to ensure that compensatory measures were properly implemented as required.- Personnel access was. observe aI to be controlled in accordance with established procedures. The PAL

- barrier had ' adequate illumination and the isolation zones were free -j of transient material The inspectors performed periodic tours of the reactor plant to .

verify proper system lineups and cleanliness. - -The inspectors verified periodically that electrical lineups were maintained for components needed to mitigate.an accident. The inspectors determined that

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i housekeeping throughout the plant had.been maintained at'an excellent level throughout the inspection period. The inspectors performed a detailed walkdown of the control. rod drive' system. ' Results of this '

walkdown are documented in-paragraph 7 of this repor The licensee reduced power on July 6,1990, to adjust the balance on-a low pressure turbine bearing. The bearing had experienced ~ a cteady j rise in measured vibration .since the plant startup in May 1990; JThe-highest observed vibration level was 6.3 mils-which exceeded the i-alarm setpoint (6.0 mils). No actions were required at the alarm -!

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The inspector observed licensed operators decrease power from 32-27 percent thermal power. The operators halted the power reduction- ;

due to a failure of the rod sequence control system (RSCS). The RSCS '

enforces a predetermined' control rod insert / withdraw sequence' below 20 percent power to minimize the consequences of a rod drop acciden The RSCS will cause rod block! to occur whenever the control rods' are moved out of sequenc .

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The RSCS surveillance procedure, which verifies system operabil_ity,  !

could not be completed. Investigation by Instrumentation and~ '

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Control (I&C) technicians identified a bad relay and a. bad toggle switch. The toggle switch is used during the conduct of the' 4 functional test. The failure of the RSCS functional test did not involve a safety problem because the control rods could still scram -l if required. The licensee returned to 100 percent power without adjusting the bearing balance because repairs to the RSCS could not -

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be accomplished in the schedu' led tim Monitoring of the bearing vibt.so. after the return to full power indicated that the vibrath, nad held steady at approximately i

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5.0 mils. The licensee is developing a method to test the RSCS to l

determine' operability at higher power levels, in' order to. prevent similar- problems in the future.- During backwashing .ithe main condenser on July'14,1990. a ' moto operated circulating water:(CW) valve bound.up and-became stuck in =

the'open position. -The s+uck,open valve' reduced flow through the, 4

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condenser causing a decrease in: vacuum. Operators reduced power to

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approximately 650 MWe. iThe CW val _ve was manually closed until . ,

backwashing was completed.1 Operators manually. re' sened the CW. valve - c'

, and subtequently returned the reactor to 100 pert it power. The _

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licensee: suspected that-the CW valve problem is due to problems with the motor-operator; limit switche ~

1 On June 28',1990, the onsite safety review'connittee reviewe'd the 10 CFR 50.59 safc+y evaluation, which justified _ increasing the-allowable river water temperature from 85-90*F. The safety evaluation q supported a USAR change to the service water inlet temperature  !

-licensing basis. Affected plant: procedures:were modified accordinglyt ]

In anticipation _of potentially higher? river temperatures, the nuclear ,

engineering'departmenttinvestigated and . determined that- the licensing basis temperatures of service water and reactor equipment-cooling ~ could

'be, raised above 90 and' 95'FFrespectively0 The' licensee drafted justification ior continued operation:(JCO) in the _ event it-would- be" q needed. -The onsite review connittee myiewed the JC0 for technical,  ;

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adequacy but was waiting!until a need arose:before' formally approving '

.the JC During a plan nonstandard tou door on'. June

tops 27,19907 on two fire doorsthe inspector located identified in the'radwaste i corridor. These non-Appendix R fire doors have .a 3/4-hour. fire - ,

rating and provide personnel. safety.. From record reviews the- 1 inspector. determined that the doors hadl the proper fire ratings, but '

the doors had not been-tested with allfappurtenancesland did notlhave analyses supportinr the addition of door stops. ' The licensee a conducted analyses to verify adequacy for. the two . doors. The licensee . l will determine whether anyf other fire doors have similar deficiencies- I and provide the analyses lresults by September 15,.1990.-

No violations or deviations were identified in this area. The licensee approved a USAR change prior te elevated river temperatures occurrin The licensee further evaluated the effects' of ttmperatures: greater than 90'F and had available the basis for continued operation under these a conditions.:

i 4. ' Monthly' Surveill ance ' 0bservations (61726) j s . ,

a., On-June 19, 1990, the inspector observed chemistry technicians 1 perfonning a field walkdown verification-of drait Chemistry 1 Procedure 8.4.1.1, " Post Accident Sampling System." The inspector- 'l e

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< observeo the chemistry technician;take a liquid sample ~and subsequently' blow down the system with nitrogen... The chemistry technicians verified that the soon to be issued procedure revision,-

which Hil follow the= licensee's Writer's Guide, properly accomplished:

the requid tasks. The technicians were knowledgeable about- the (

purpose of the cystem. A copy of- the existing, approved procedure-tes atbched to the draft in the event a question arose. regarding operation of Ine system, b. - On' June 20, 1990, the inspectors observed the performance of

- Surveillance Procedure (SP). 6.3.6.1, "RCIC Test Mode. Surveillance -

Operation " Revision 24 dated April. 23, 1990. This test-was-

. performed as a monthly-functional test' to verify compliance with-Technical Specification (TS) operability requirements.. A. station operator-in an a) proved license class operated the control' panel-switches under tie guidance of a licensed operator. - The licensed opcator provided instruction to the trainee on expected controlf panel indications and system response as the test progresse The inspector observed a station operator take monthly ' vibration readings at the reactor core isolation cooling (RCIC) turbine. The station operator correctly swapped-between the magnetic and stick probes used to take vibration measurements.

'- On June 20, 1990, the inspector observed the performance-of j SP 6.3.20.1," RHR Service Water Booster Pump Flow Test' and Valve- l Operability Test," Revision .24, dated April 30, 199 This test was performed to verify operability per TS and obtain quarterly. inservice test data. The station operator used calibrated I test instruments. The test data, for the pump observed, met .

specifications. Review of the completed procedure indicated that the l measured: vibration on' Service Water Booster Pum) (SWBP)'"A"'placed it j in the alert range. . The. inspector determined tlat previous. tests had placed SWBP "A" .in the alert range and that it was scheduled'to be tested in accordance with' increased frequency requirement d., On June 21, 1990, the inspector observed the performance of'

SP 6.2.2.2.1,." ADS Water Level Calibration and Functional / Functional u Test," Revision 23; dated April 16, 1990. This. test was performed'as D a :nonthly functional test to verify operability as required by T The, test verified that the reactor water level' indicating switches actuated at the propcr level to initiate the automatic depressurization system timer permissive signal and that the necessary relays actuate The inspector noted that proper radiological practices were followe The technicians bagged test equipment taken into the contaminated area. Test hoses routed from.the-cleen area'to the contaminated area had been sleeved to prevent contaminating the hoses, 'All data met specification ;

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' On Jurce 20, 1990, the inspewtors observed SP 6.2.2.1.3, "CSCS Reactor:

Low Pressure Valve-Permissive Calibration and' Functional / Functional-

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. Test," Revision 19, dated April 30, 1990, beino performed. : This' test . 4

. verified that the core. standby cooling _ system (CSCS) injection valves got a permissive _ to open and the reactor. recirculation discharge f; valves got a permissive to.close on decreasing reactor. pressur '.

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This' functional test performance verified instrument operability as . ,

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required by TS. The inspector noted:that_ the relays, actuated as -

specified in the procedur '

f.- On June 22, 1990, the. inspector observed operators perform '

SP 6.2.2.1.10, "4160 V Buses 1F and 1G Undervoltage Relays and Relaj l

- Timers Functional _ Test,". Revision 16, dated December: 28,1989.._ This j test was performed to verify operability-as required-by: TS. .The - ,

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. licensed operating crew performing the test.had not recently performed 1

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this particular surveillance. The -licensed operators took~ extra ,

precautions prior to step perf mnance te ensure a safe ~ test e pu formance and reviewed / discussed appl' cable procedure' steps prior-to performing each portion of the test. Excellent communications were maintained between the control room.and the test performer ;

One time-delay relay was determined to actuate octside the required time limit. A maintenance work request (MWR) was Lwritten, the ~ problem L corrected, and a retest of- the time-delay relay _ performed ,;

satisfactorily (see paragraph 5). The operators were aware of the--

precautions and limitations of this procedur ,

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r On July 3, 1990, the inspector observed the performance of

, SP 6.2.1.2.1, "PCIS RWCU High Flow Calibration and Functional /

Functional Test," Revision 14, dated April 23 -1990. 'The I&C- o

technicians performed this test to verify' operability. as required by '

i TS. The test verified that the primary containment isolation system- ,

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actuated during high flow conditions in the reactor water cleanup

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The inspector observed'the calibration of a Barton differential .

. pressure indicating switch,-RWCU-DPIS-170B. The instrument required _ ,

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calibration track the test because values the linearity).(the-ability being of'the instrument input was,out of specification even to though-the trip arm remained at the proper setting. _ When the

technician started to adjust the linearity, as allowed by the SP, a '

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screw broke and the technicians secured the test af ter assuring' that

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the instrument tripped at the proper. setpoint. After correcting the problem (see paragraph 5), the technician completed the calibration of the instrument.

] Discussions with I&C personnel and licensee management revealed j questions about whether the technician performing adjustments on the .

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instrument had received proper training on the instruments. The l 1- issue of proper' training for I&C'techniciens to perform instrument' 4

adjustments is considered an unresolved item (298/9026-01) pendin the licensee providing training records'to the inspector ;

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l L 8 No violations or deviations were identified in this area. Exceller,t communications were observed during all tests, in that, test performers were' aware of the steps being performed and had copies of the test i procedure to minimize potential confusion. Good i sdiological practices were demonstrate . Monthly Maintenance Observation (62703) On July 5, 1990, the inspector reviewed maintenance activities related to setting the time delay on an Agastat relay (an i electropneumatic device). Electricians had set the time delay at 7.2 seconds from a measured reading of 8.2' seconds since the maximum .

time delay for this relay was 8.0 seconds. This had been the second- l'

incidence of this particular Agastat relay drifting. In both cases a nonconformance report (NCR) had been written. Discussions with the ;

responsible electrical system engineer indicated that the corrective actions response to the first NCR was to reset the time delay and -

monitor the relay-perfomance. The response to the second NCR will be a recomendation to replace the Agastat relay with a solid state relay, On July 5,1990, the inspector reviewed corrective maintenance activities for RWCU-DPIS-1708. This particular Barton DPIS had dual alarm setpoints. The alarm setpoints are adjusted by lef t-hand and j right-hand adjustment arms. During conduct of a surveillance on '

July 3, a screw used to adjust the linearity of the indicating switch c had broke The I&C technician researched the vendor manual and determined that the broken adjustment screw, which went to the right-hand trip arm, was not needed because the instrument uses the left-hand trip settin ;

Additional discussions with the responsible electrical system engineer confimed that the left-hand trip arm was used in this applicatio The-technicians removed the internal linkages for the right-hand trip ;

arm to eliminate any potential for future interference. The '

technicians subsequently completed the instrument calibratio No violations or deviations were identified in this area, t Onsite Followuo of Written Reports (92700)

(Closed) LER 88-019: Manual Reactor Scram and Subsequent Engineered Safety-Feature System Actuation Due to Isolated Phase Bus Duct Grounding Proble The licensee reduced power from 100-50 percent by reducing recirculation flow and scramed the reactor due to arcing at the bolted connection of an isolated phase bus duct cover. The licensee experienced Group 2, 3, and 6 isolations due to expected reactor vessel level shrink, even though vessel level was raised in anticipation of this shrin I l

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The licensee determined that the arcing was caused by failure of the-insulating gasket on the removable cover of the phase bus duct. The '

gasket was replaced, ground straps and bolting damaged due to arcing were replaced, and the other isolated phase bus ducts and buses were inspecte t The licensee initiated a preventative maintenance item to periodically inspect and/or replace the insulating gaskets ~on the removable-phase bus duct covers. This LER is close (Closed) LER 89-001: Unplanned Automatic Scram Due to APRM High Flux >

Resulting from Separation of an MSIV Dist from its Ste The disc of a main steam isolation valve (3SIV) separated fran the stem i and isolated one steam line with the plant' at 1C0 percent power. This event is analyzed in the plant design docunintation._ The event followed predictions, with the resulting reactor pressure. spike causing a high neutron flux scram. - Plant systems functioned as designe The-licensee, with assistance from the reactor vendor, concluded that the separation of disc and stem in this kind of valve is a random event. The

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disc and stem of the failed valve were replaced with a newly. designed disc and stem assembly intended to prevent future failures. ' While the steam line was' isolated, the remaining MSIV on this-.line was fitted with the new design stem and disc assembly. The remaining six MSIVs were

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equipped with the new stem and disc assemblies during a subsequent refueling outage. This LER is close (Closed)LER89-002: Unplanned Actuation of the Reactor Protection System and Engineered Safety Feature Group Isolations Due to Inadequate-Job Planning While Shutdow .

With the plant shut down, the licensee attempted to measure; the differential pressure between the sensing lines to the reactor vessel level sensors. When the differential pressure' test instrument was valved in, the reference leg of the level gauge filled with water, generating a

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reactor scram signal due to trippiag of a lavel switch from each reactor protection ' logic train. The switches used the same reference leg. The licensee concluded that the I&C technicians should have foreseen the scram resulting from this test and should have conducted the test in a way tha+ did not result in a scram. 1&C personnel have been retrained regarding this event. Planning tests to minimize plant impott has been reemphasized. This LER is close (Closed)LER89-004: Inadvertent Actuation of No.1 Emergency Diesel Generator (EDG) Overspeed Trip With the Engine in Standby While No. 2 Emergency Diesel Generator was Inoperable for Maintenanc ~

l Licensee personnel bumped the Safety Shutdown Valve (SSV) overspeed trip

' lever on EDG No. I while taking measurements on control air tubing. The plant was operating and EDG No. 2 was out of service. EDG No.~1 was made operable within 2 minutes of beins : ado' inoperable by bumping the SSV overspeed trip lever. The licensee has posted the access to the walkways

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on the EDGs requiring ' permission from the-shift supervisor to gain acces Additionally, a permanent guard plate was installed'around the SSV  !

overspeed trip. lever. This LER is close ,

(Closed) LER 89-016: Unplanned Automatic Initiation of an Engineered-Safety Feature While Conducting a Surveillance Test Due to a Human Factors Deficiency.- ,

With the reactor shut down and depressurized for refueling, core spray-was actuated during performance of a surveillance test. Core spray was-

- immediately terminated by the. control room operator after injection of 500 gallons. The licensee determined that this had no safety impact. A technician installing relay contact blocks, to prevent this actuation, had counted from-left to right when identifying the relay contacts to be ,

blocked. The. relay was marked opposite frcm the side faced by the ,

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technician, consequently, on the side the technician faced, the contacts -

were numbered from right to left. The licensee reviewed:all surveillance procedures and identified those procedures requiring installation'of contact blocks or jumpers. Procedures requiring. installation of jumpers :

- or contact blocks were revised to cluify the installation instructions to prevent similar events. . This LER is close (Closed)'LER-89-020: Unplanned Group Isolations During D(sign Change t Acceptance Testing Due to Wiring Error '

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During testing of a design charige associated with the Division.1 diesel

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generator start circuitry, power to a safety-related 4160 V bus was los This resulted in-several group isolations. The licensee determined that ,

the switch designed to prevent deenergizing this bus had been wired in'

reverse to the way indicated on existing drawings. = This reversal-had no effect at the time, and the switch worked until the current' design change' ,

l was performed. With the actml lads reversed from what the ~ drawing showed, the current design chenge took the switch out'of the test circuit

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and allowed-the bus.to be-deenergize >

The licensee had an as-built drawing ver cation program in progress-.at -

the time of the event. The particular sings involved had not yet been verified. The licensee issued instrue' ans that no drawings were to be .

used for design changes until the dra'..ngs were field-verified. This'LER is closed.

! (Closed)LER89-011: Failure to Identify Unsatisfactory Test Results Which Resulted in Exceeding . Technical Specification Surveillance Requirements for an Inoperable Residual Heat Removal Pum A . test performed at low flow conditions produced results uutside the acceptance criteria on an RHR pump. The shift supervisor ~ failed to .

identify'the unacceptable results. Consequently, the r:mp was not. declared ,

inoperable until more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after receipt of the unacceptable test results. The-pump was subsequently tested and found to be operable.

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The licensee's system worked as the second of three reviews of test

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results identified the unacceptable test result. The shift supervisors 1-were retrained regarding their responsibility for test result revie This LER is closed.- '

No violations or deviations were identified in this are ., ESF Walkdown- (71710) 1 The inspector conducted an independent verification of the control rod drive (CRD) system status. The inspector compared the walkdown sheets

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from Station Operating Procedure 2.2.8, " Control Rod Drive System,"

Revision 21',-Appendix A, dated July,27, 1989, to the latest controlled '

drawing.. Subsequently, the inspector walked down accessible portions of_ ,

the system, using the same checklist. The drawing had a few minor errors, ;

such as valve position errors and missing vent and drain vtives and, "

during the field walkdown, one tag was identified as missing. The inspector notified the licensee of these discrepancies. The licensee responded that the CR0 system was' included-in the scope of the as-built j program and that the drawing discrepancies would be corrected at the time

- of the system walkdow All instruments were valved in and functioning properly. Local flow indication agreed with control room indications. Power was available to valves as appropriate.

r No violations or deviations were identified in this area.

' Inoffice Review of Event Reports (9u/12I Inoffice review of licensee event reports (LERs) was performed to verify the following: l

. Correspondence included the information required by appropriate NRC requirements;

. Correspondence did not contain incorrect, inadequate, or incomplete informat!on;  ;

. Evaluations were perfonned to determine safety significance of the event;

. The identified root cause of the event is-supported by:the data provided; and

. Planned corrective actions were adequate for resolution of identified i problem Twenty LERs were reviewed during this inspection period and are listed in Attachment "A." These reports encompassed all of the LERs that the licensee had submitted between April 7, 1989, and July 5, 199 ,

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l' It was determined that all of the LERs reviewed met the criteria listed abov In addition, a review of the root causes of-the events described in each LER identified 'no trends. or programmatic concerns with the manner in which the licensee had operated the plan No violations or deviations were' identified in this are ; Unresolved Item An unresolved item is one about=which additional information is required in order to determine if it-is acceptable, a deviation, or a violatio There is or's unresolved item in this repor Paragraph Item N Subject ,

4.g- 298/9026-01 Adequacy of I&C Technician Training 10 - Interviews ' (30703)

An exit interview was conducted on July. 17,'1990, with-licensee representatives identified in paragraph 1. .During _ the interview, the inspectors reviewed.the scope and findings of the inspection. Other meetings'between the inspectors and licensee management were held periodically during the inspection period to discuss ' identified concern ~

The licensee did'not identify as -proprietary any information provided to, .

- or: reviewed by, the inspectors. -

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ATTACHMENT ,

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LER' COMENTS EVENT DATE

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.89-012- Actuation of Engineered Safety . 04107-90 FeaturesL(ESF)GroupIsolationSubsequent,

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to a Planned Manual Scram Due to Momentary i Low. Reactor _ Vessel Water Level ,89-013 Unplanned Actuation ~of ESF Feature Group 04-14-89-

~Isolations While Replacing. Relay Coils-89-014 Unplanned Acttiation of Group 6 Isolation 04-24-8 Duezto a Loose.TerminalzBoard Connection Found.While Performing. Design: Change and ,,'

l Maintenance Actifities'89-015 Safety / Relief Valve Setpoint' Variance;Not 05-03-89 t Within Technical Specification Limit- T

,89-016 Unplanned Automatic Initiation of an .05-10-89 ESF While Conducting a. Surveillance Test Due'to a Human Factors Deficiency

,89-017 Failure of Standby Liquid Control 05-11-89 System Squib Valve to Function During Surveillance Testing Due to Primer '

^5 amber Pins BeingiShunte ~

89-018 L planned Scram and Containment ~ 05-18-89

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Isolations While Shutdown.Due1 toia . i Loss of Power to One Trip ~ System While  ;

Testing was in Progress:on the Opposite'

Trip System 89-019 Inadvertent Actuation of Group Isolations 05-24-89 While Perfonning Design-Change Activities

.Due.to Lifting Incorrect Leads89-020 Unplanned Group -Isolations-During Design 05-29-89 Change Acceptance -Testing Due to Wiring

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Errors-'89-021 Undocumented Wiring Configurations 06-02-89 Associated with Safety-Related Equipment Discovered During Design Change Activities89-022 Identification of a Condition Which Could 06-13-89 Have Rendered Both Trains of Standby Gas Treatment Inoperable 1

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.LE COMENTS'. EVENT DATE: , .,

,89-023 Valve, Body Wall-Thin 61ng in Safety 05-22-89- -i

' elated Throttle Valves Due to Erosion'89-024 Diesel Generator-Injection Pump .

=07-31-89 Pedestal' Crack Which was Caused.by-Unusuai~ Hydraulic Forces Found DuringL

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Troubleshooting Subsequent to- '

Surveillance Testing

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89-025- Unplanned Main Turbine Trip and' 09-28-89 Subsequent Reactor Scram Caused by a Spurious Main Turbine Hydraulic Control Oil Reservoir-Low Level Signal 89-026 Reactor Scram Due to Main: Steam Isolation '11-25-89'- -!

Valve Closure as a Result of Low Instrument Air Pressure Caused by - i

,an Air Dryer Malfunction 90-001 Unplanned Isolation of.the High' Pressure 01-24-90 Coolant Injection System.During Surveillance Testing.Due to Human Factors and Procedural Deficiencies ,90-002 Actuation;of ESF Group Isolations- 03-03-90 Subsequent to:a Planned Manual Scram and During Cooldown Due to Design

Characteristics and Required Trip System Settings

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s90-003 -Safety / Relief Valve and Safety Valve' 03'-28-90 Setpoint Variance Not Within Technical Specification Limits ,

.90-004 ESF Group Isolations and Diesel Generator 04-13-90 Starts Due to-Equipment Malfunction and- ,

Personnel Error 90-005 Unplanned Actuation of-Group Isolation 04-30-90 ESF Upon Loss of Power From the B RPS Motor Generator Set Due to a Relay ,

Contact-Deficiency

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