IR 05000254/1987033

Insp Repts 50-254/87-33 & 50-265/87-33 on 871206-880206. Violations & Deviations Noted.Major Areas Inspected: Operations,Maint & Surveillance,Ler Review,Routine Repts & Administrative Controls Affecting Radiation Control

ML20149N087
Person / Time
Site:	Quad Cities
Issue date:	02/24/1988
From:	Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20149N081	List: IR 05000254/1987033 ML20149N080 ML20149N084 ML20149N091
References
50-254-87-33, 50-265-87-33, IEB-85-003, IEB-85-3, IEB-87-002, IEB-87-2, NUDOCS 8803010168
Download: ML20149N087 (14)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-254/87033(DRP);50-265/87033(DRP)

Docket Nos. 50-254, 50-265 Licenses No. DPR-29; DPR-30 i

Licensee: Commonwealth Edi.va Company Post Office Box 767 Chicago, IL 60690 Facility Name: Quad Cities Nuclear Power Station, Units 1 and 2 Inspection At: Quad Cities Site Cordova, IL Inspection Conducted: December 6, 1987 through February 5, 1988 Inspectors: R. L. Higgins

A. D. Morrongiello Approved By: M. A. Ring, Chie N E Projects Section IB f/28T

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Inspection Summary a

_ Inspection on December 6, 1987 through February 6, 1988 (Reports N /870033(DRP); 50-265/87033(DRP))

i Areas Inspected: Routine, unannounced resident inspection of Operations, i Maintenance, surveillance, LER Review, Routine Reports. Administrative  ;

Controls 40'scting Quality, Radiation Control, and Outage '

Resul ts: U, the creas inspected, two deviations, one violation, and one

licensee twentified violation were identifie I i

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8803010168 PDR 8poca5 ADOCK Oa000254 0 DCD '

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! Personnnel Contacted (

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• T. Tamlyn, Production 3 qcriniendent  ;

• M. Kooi, Regulatory-Assurance Supervisor  ;

l *A. Scott, Quality Assurance Manager- l r

• Denotes those present at the exit interview on February 8,1988,

The inspectors also contacted and interviewed other licensee and contractor personnel during the course of thi: inspectio ; Operations (71707,93702)

l The inspectors, through direct observation, discussions with licensee l personnel, and review of applicable records and logs, examined plant operations. The inspectors verified that activities were accomplished in 6 timely manner using approved procedures and drawings end were !

inspected / reviewed as applicable; procedures, procedure revisions and j routine reports were in accardance with Technical Specifications, regulatory guides, and industry codes or standards; approvals were i obtained prior to initiating any work; activities were accomplisned by r qualified personnel; the limiting conditions for operation were met ;

during nonnal operation and while components or systems were removed from i service; functional testing and/or calibrations were performed prior to l'

l returnin3 components or systems to service; independent verification of equipment lineup and review of test results were accomplished; quality l control records were properly maintained and reviewed; parts, materials !

and equipment were properly certified, calibrated, stor d. and or (

maintained as applicable; and adverse plant conditions ti.:'oding !

equipment malfunctions, potential fire hazards, radiological hazards, ,

fluid leaks, excessive vibrations, and personnel errors were addressed l'

in a timely manner with sufficient and proper corrective actions and reviewed by appropriate manage w.; personne (a) Engineered Safety Features System Walkdown (71710)

I During plant tours of Units 1 anil 2, the inspectors walked down the !

accessible portions of the High Pressure Coolant Injection Systems, i Reactor Core Isolation Cooling Systems, Core Spray Systems, Residual l Heat Removal Systems, Standby Liquid Control Systems, Standby Gas !

Treatment Systems, and Diesel Generator i (b) Sumary of Operations j l

Unit 1 l At the beginning of the inspection period Unit I was in a scheduled refuel outage. At 6:06 PM CST on 12/22/87 control rods were I withdrawn to start up the reactor and at 10:00 PM CST on 12/22/87 1

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the reactor achieved criticality. At 5:25 PM CST on 12/23/87 the generator was synchronized to the electrical grid, but at 5:50 PM l CST the turbine was tripped due to high vibrations, with the reactor remaining critical in hot standby. The generator was synchronized to the electrical grid at 1:20 AM CST on 12/24/87, but high ,

vibrations necessitated the generator be removed from the electrical grid at 2:32 AM CST and the turbine tripped at 3:55 AM CST, with the reactor remaining critical in hot standby. The generator was synchronized to the electrical grid again at 1:10 PM CST on 12/24/87.

l At 11:05 AM CST on 12/27/87 the turbine was tripped as part of an overspeed test; the reactor remained critical in hot standby.

, Problems with the turbine control prevented the completion of the

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turbine overspeed test until repairs were made on 12/29/87. At 5:03 l PM CST on that date the generator was reconnected to the grid. The

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remainder of the overspeed test was completed at 8:40 PM CST st

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which time the turbine trippe The reactor remained critical in hot standby until 10:10 PM CST on 12/29/87, at which time the generator was reconnected to the electrical grid and power increased to rated loa At 8:50 Pti CST on 1/2/88 an electro hydraulic control fluid leak l was discovered on the front standard of the high pressure turbine, necessitating load to be reduced and the turbine tripped at 1:35 AM CST on 1/3/88. The reactor remained critical in hot standby while repairs were made to the electro hydraulic control system. The generator was connected to the grid at 8:57 AM CST on 1/3/S i For the remainder of the inspection period Unit 1 operated either at full power, on Economic Generation Control (EGC), or at reduced -

power in order to perform surveillance testing or to respond to load dispatcher orders. As of the end of the inspection period l the unit had operated at power for 33 consecutive day Unit 2 During the inspection period, the unit operated ei+her at full power, on Economic Generation Control (EGC), or at reduced power in order to perform surveillances or to comply with load dispatcher orders, until the reactor sc.ramed from 100% power at 1:17 AM CST on 12/10/87 due to a failed Master Trip Solenoid causing a turbine generator load reject signa At 7:50 AM CST on 12/10/87 rod withdrawal began to restart Unit 2 and at 11:41 AM CST Unit 2 attained criticality. At 9:07 PM CST the Master Trip Solenoid was replaced and passed an acceptance test. At 12:30 AM CST on 12/11/87 the main generator was connected to the ;

electrical gri '

Unit 2 continued to operate either at full power, on EGC, er at reduced power to perform surveillance testing or comply with load 3 I

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dispatcher orders, until 10:55 AM CST on 1/11/88, when the unit scrammed from 100% power due to a ground on the main generator or main transformer. Investigation by the licensee could not locate the cause for the' ground bmse the ground disappeared soon after the unit scrammeo. At 9:00 AM CST on 1/15/88 control rods were withdrawn to restart the re actor, and criticality was attained at'

10:57 AM. The main generator was cconected.to the electrical grid at 3:00 AM CST on 1/16/8 For the remainder of the report period, the unit was either at full power, on Economic Generation Control (EGC), or.at reduced power in order to perform surveillances or to comply with load dispatcher orders. As ci the end of the inspection period the unit had operated at power for 22 consecutive day (c) Degraded Voltage Relay Timer Exceeds Technical Specification Limits At 1:30 PM CST on 12/07/87, with Unit 1 in cold shutdown in the midst of a refuel outage and Unit 2 on Economic Generation Control above 90% power, a surveillance test performed on the degraded voltage relay timer for the Unit 14160 volt essential bus 13-1 indicated that the timer exceeded Technical Specification limit Technical Specification Table 3.2-2 requires the degraded voltage timers for the 4160 volt essential buses to insert a time delay of between 285 and 315 seconds. The actual time delay displayed by the timer for bus 31-1 was 319 seconds. The time delay was immediately reset to a value between 285 and 315 seconds. The degraded voltage time delay for the other Unit 14160 volt essential bus was found tc be within Technical Specification limit (d) Reactor Scram At 11:50 PM CST on 12/09/87, weekly turbine generator tests were being performed on Unit 2. The Master Trip Solenoid Test switch was placed in the "Trip A" position and the "Trip A" light extinguished as required. The Master Trip Solenoid Test switch was returned to the "Reset" position, but the "Trip A" light failed to illum aate as required by the test procedure. The test was suspended. At 12:50 AM CST on 12/10/87, the "Trip B" light also extinguished, though the Master Trip Solenoid Test switch had not been manipulate Instrument technicians were notified and began investigating. At 1:17 AM CST on 12/10/87, Unit 2 scrammed on a turbine generator load reject signal. At the time of the scram Unit 2 was at 100% power and Unit 1 was in cold shutdown in the midst of a refuel outag ,

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Refer to paragraph 3.(b). '

The only abnormalities observed during the scram were rod G-11 drifting out to the 02 position after fully inserting to the 00 position, and the B feed water regulating valve locking up in the i fully open position. Rod G-11 was manually inserted to the 00 '

position after the scram was reset. The B feed water regulating valve locking up in the fully open position caused reactor vessel I

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. water level to rise above +48 inches, resulting in the feed water pumps tripping off. The reactor water clean up. system was.used to reduce reactor level. The B feedwater pump was restarted, but th !

A feedwater regulating valve leaked excessively, resulting in reactor vessel water level again rising above +48 inches, causing :

thr B feedwater pump to trip. The reactor water cleanup system was again used to restore reactor vessel water level .to' normal, the A feed water regulating valve was. isolated,-and the B feed water pump was restarted and used t7 maintain a stable reactor vessel water level. The abnormalitiu mere properly addressed on 12/10/87, and rods were withdrawn to restart the reactor that same da (e) Supports for Instrument Racks Do Not Meet ASME Requirements At 7:50 PM CST on 12/19/87~ the Quad Cities plant personnel were notified by the Commonwealth Edison corporate BWR engineering grou that the ATWS support hangars for the reactor vessel instrument racks on Unit 1 (racks 22015 and 6) and Unit 2 (racks 2202-5 and 6) '

'do not meet ASME Code. requirements for seismic support but do meet operability requirements. .At the time of the notification Unit 1 was in the 98th day of a refueling outage and Unit' 2.was at 100%

power. Failure to comply with ASME requirements is considered to be a Deviation (254/87033-01(DRP) and 265/87033-01(DRP)) from the licensee's commitment to the ASME code, h discrepancy was discovered by the Senior Resident .nspector when touring the plant with the licensee's Assistant Superintendent for Operations on 12/1/87, during which a loose support was notice The Assistant Superintendent for Operations wrote a work request documenting the loose support. Further investigation by engineering led to the determination that the supports for the instrument racks did not meet ASME requirements for seismic support, but did meet operability requirements. The licensee mcdified the supports for Unit 1 so that they met ASME requirements prior to startup, and will modify the supports for Unit 2 during its next refueling outage scheduled to begin in April 198 Further investigation by the' licensee revealed that the instrument racks were not installed according to the design procedures and drawings. This is considered to be a violation of 10 CFR 50 Appendix B. Criterion V., which requires activities affecting quality to be prescribed by documented instructions, procedures, or drawings, and to be accomplished in t.ccordance with these instructions, procedures or drawings. Because this violation satisifies the guidelines of 10 CFR 2 Appendix C, V G, Paragraph 1, no Notice of Violation will be issued (254/87033-02(DRP) and (265/87033-02(DRP)).

(f) Unit 1 HFCI and RCIC Abnormalities At 1:10 PM CST on 12/23/87 a valve operability test was being conducted on the steam supply valve for the Unit 1 HPCI pump. When the steam supply valve was opened the minimum fled valve from the l

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HPCI discharge lina back to the torus should have opened but.did not. This rendered HPCI inoperable. At the time of the valve operability test Unit I was at 13% power in the process of starting up froc a 102 day refueling outage with the Unit 1 RCIC turbine uncouraed from its pump while undergoing overspeed tests. Thus both HPCI #.nd RCIC were inoperable at the same tim In accordance with Techical Specification 3.5.C.3 this condition requires the reactor be shut down and reactor pressure be reduced below 90 psig within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The RCIC overspeed test was successfully completed and the RCIC turbine ru:oupled to the RCIC pump in preparation.for its operability test. The problem with the HPCI minimum flow valve was determined to t,e a faulty low flow relay which was replace At 4:53 PM CST on 12/23/87 a manual initiation test of RCIC was performed. The test was satisfactory except that RCIC did not inject into the reactor vessel. RCIC was therefore declared inoperabl At 8:40 PM CST on 12/23/87, after the HPCI minimum flow -valve was -

repaired, the HPCI quarterly surveillance test and valve operability test were performed satisfactorily. Unit I was no longer bound by the requirement to shut the reactor down and reduce pressure below 90 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but now tmcame ' subject to Technical Specification 3.5.E.2, which allows reactor operation for 7 days, at which time the reactor must be shut down and pressure reduced below 90 psig within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The reason that RCIC did not inject into the reactor vessel was detemined to be a check valve in the line to the reactor vessel being stuck shut. It is thought that this check valve became shut !

during the hydrostatic test of the reactor vessel conducted during the outage. The check valve was manually agitated in order to free ;

it, and RCIC was subsequently retested 15 times, each time !

sucessfully injecting into the reactor vessel. Unfortunately, the l RCIC did not consistently start within the 30 seconds required by the FSAR, and the RCIC minimum flow bypass valve did not function properly. RCIC therefore remained inoperabl At 8s25 PM CST on 12/24/87 the Unit 1 HPCI pump was undergoing an operability test because RCIC was inoperable. In order to perform !

the operability test on HPCI, the normally-open HPCI pump discharge

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valve 1-2301-9 had to be shut so that the operability of the normally-shut HPCI pump discharge valve 1-2301-8, which must open I automatically when HPCI starts in order for HPCI to inject into the reactor vessel, could be teste The 1-2301-9 valve could not be shut from the control room, so the HPCI operability test could not be performed. HPCI was therefore declared inoperable. Since RCIC was already inoperable Technical Specification 3.5.C.3 appliert, which requires that the reactor be shut down and reactor pressu*e reduced below 90 psig within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> \

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The 1-2301-9 valve was thought to require more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to repair, and since RCIC would also_ require more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to repair, licensee management decided.to shutdown Unit 1 at 9:55 PM CST on 12/24/87. In accordance with Quad Cities Emergency Plan Procedures, which require-an Unusual Event'to be declared when a. Technical Specification LC0 necessitates-a: reactor shutdown, an Unusual Event was declared at 10:00 PM CST. The Illinois Emergency Services and Disaster Administration. (ESDA) was notified at 10:03 PM CST ~and the NRC Emergency Operations Center was notified at 10:06 PM CST. At 10:15 PM CST Unit 1 began shutting dow Investigation revealed that the 112301-9 valve could be shut locally, so the operability test on the 1-2301-8 valve could be perfonned, and therefore the HPCI operability test could be conducted. Since the 1-2301-9 valve is a normally-open valve which has no lutomatic function, it was decided that HPCI would not be declared inoperable simply because the.1-2301-9 valve could not be shut from the control room. The HPCI operability test was conducted at 2:20 AM CST on 12/25/87, and except for the automatic flow controller controlling flow 500 gpm above ,its setpoint, the test was flawless. The excessive flow provided by the flow controller was not deemed to detracc from HPCI operability. The Unusual Event was terminated at that time; ESDA was notified at 2:27 AM CST and the

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NRC Emergency Operations Center was notified at 2:55 AM CST on 12/25/8 With HPCI returned to service, Unit I remained subject to Technical

' Specification 3.5.E.2, which requires that RCIC be returned to service within 7 days or the reactor be shutdown and reactor pressure reduced below 90 psig within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The Senior Resident Inspector was in the control room and monitored the HPCI operability tes The RCIC discharge check valve was disassembled and repaired, and the RCIC fast-stsrt rush button and starting mechanism were repaired. RCIC wa; successfully tested, including reactor vessel injection, and returned to service at 8:53 AM CST on 12/28/87 (

Because of the numerous problems with RCIC and HPCI evident during the startup from the Unit 1 outage, additional attention is being devoted to the RCIC and HPCI systems by the Quad Cities Resident Inspectors and by Region III personne (g) Inadvertent Control Rod Scram into Unit 1 Reactor On 12/26/87 Unit I was at 24". power conducting hot scram timing tests. At 1048 AM CST the hot scram timing test for rod 30-35 was performed using its scram test switch. The scram test switch for rod 26-39 was inadvertently manipulated at the same time, carsing

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rod 26-39 to scram into the reactor from its fully-withdrawn ;

position of step 48 along with rod-30-35. Because reactor power

exceeded 20%, the Banked Position Withdrawal Sequence was not a concern. At 1052 AM CST rod 26-39 was withdrawn to step 48. At 11:20 AM CST the NRC Emergency Operations Center was notifie (h) Unit 2 Reactor Scram At 1055 AM. CST on 1/11/88 Unit 2 scrammed from 100% due;to a ground on the main generator. All safety systems functioned properly during and after:the scram. The SRI was in the control room -

innediately after the scram to monitor post scram activities. The ,

NRC Emergency Operations Center was notified of the event at '

1120 AM CS Extensive investigation by the licensee could not locat'e the cause-for the ground because the ground disappeared soon after Unit 2-scranned. At 9:00 AM CST on.1/15/88 control rods were withdrawn to startup Unit 2, and criticality was attained at 10:57 AM CST. Th Senior Resident Inspector was in:the control room and monitored the approach to and attainment of criticality. Unit 2 was connected to i the electrical cid at 3:00 AM CST on 1/16/8 (i) ESF Actuation On February 1 at 0739, the control room ventilation system isolated on a Toxic Gas Analyzer - Hi Chlorine signal. At.the time instrument mechanics were refilling the chlorine monitor prob A: cording to procedure refilling the probe would result in an I

, isolation and the control room was to be notified. Contrary to i procedure the control room was not notified with the result being an unexpected ESF Actuatio '

This is considered to be a violation (254/87033-03(DRP)).

(j) ESF Actuation On February 4 at 0440 the control room ventilation system isolated on a high chlorine signal. The isolation ,ignal cleared quickly and was considered spurious. The licensee is making arrangements for the manufacturer of the probe and sampling equipment to visit the site and advise ~them on the adequacy of the present sampling installatio . Monthly Maintenance Observation (62703)

Station maintenance activities of safety related and non-safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical-Specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were

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removed from service; approvals were obtained prqu. to initiating the work; activities were accomplished using-approved procedures and were inspected as applicable; functional' testing and/or calibrations _were performed prior to returning components-or systems to service; quality control records were maintained; activities were' accomplished by qualified personnel; parts and materials used were properly certified; radiological controls-were implemented; and fire prevention procedures .

were followed. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc The following activities were observed / reviewed: Portions of magnetic particle inspection of feed water line Portions of installation of ATWS Supports on the 2201-5 and 6 Instrument Racks on Unit Portions of Unit 2 Main generator and main transformer electrical ground investigation, Portions of containment temperature detector maintenanc Portions of the investigation of the cause of the stuck Unit I control. rod G- Portions of Mechanical Maintenance installing 1A CRD pump rotor.

' Portions of Electrical Maintenance working on Unit 2 bus duct , Portions of Mechanical Maintenance repairing traveling screen '

No violations or deviations were identifie . MonthlySurveillanceObservation(61726)

The inspectors observed Technical Specifications-required surveillance testing and verified that testing was perfonned in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results' conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The following activities were observed / reviewed:  ! Portions of scram timing on Unit Portions of HPCI operabilit Portions of Control Rod Drive Temperature Detection Surveillanc I 9 ,

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- Portions of Low-Low Reactor Water Level trip and calibration surveillanc Portions of Instrument Air Compressor vibration testin Portions of Instrument Maintenance Main Steam Line Radiation Monitor Surveillanc No violations or deviations were identifie . LERReview(92700) Unit 1 (1) (0 pen) LER 87008, Revision 00: IC RHRSW Pump Piping in Excess of Allowable Stress Due to Sheared Anchor Bolts - System Still Operabl This item is being investigated by a region based inspecto (2) (Closed) LER 87021, Revision 00: 1/2 A Diesel Fire Pump Inoperable due to Intake [ Bay] Being Empty for Maintenance Activitie On November 11, 1987 at 1055 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.014275e-4 months <br />, Quad Cities Unit I and 2 were in the refuel and run modes at 0 percent and 100 percent power respectively. At this time, the 1/2 A Diesel Fire Pump exceeded the seven day reporting criteria due to being out-of-service (its intake bay was de-watered to perform maintenance on a Circulating Water Pump that shares the same bay).

On November 17, 1987, the intake bay for the 1/2 A pump was refilled. However, the 1/2 A Diesel Fire Pump continued-to be unavailable. This was due to a problem with the 1/2 B Diesel Fire Pump (re 254/87-023, Revision 00) ported in Licensee that required Event utilizing Report parts from the 1/2 A pump to repair the 1/2 8 pum This fire pump was satisfactorily testad and returned to service on November 21, 198 (3) (Closed) LER 87023, Revision 00: 1/2 B Diesel Fire Pump Starter Motor Fire Due to Loose Connection and Corrosio On November 11, 1987, Quad Cities Unit One was in the refuel mode at 0 percent power and Unit 2 was in the run mode at 95 percent tharmal power. At 5:28 AM CST, a low fire protection header pressure automatic start signal should have started the 1/2 B Diesel Fire Pum The 1/2 A pump was unavailable for maintenance reasons (p.see LER 254/87-021 Revision 00). The 1/2 !

B pump subsequently did autostart at 5:40 AM CST and the

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ElectricalMaintenanceDepartment(EMD)initiatedaninspection !

of the 1/2 B pump to determine why it did not start upon low l

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suction pressure. .The EMD observed two successful autostarts, but a third autostart resulted in the failure of the diesel engine starter motor. This occurred at 1:35 PM CST. The motor failure caused a lot of smoke and a small fire which was quickly extinguished. The 1/2 8 Fire Pump was declared inoperable. A backup fire suppression water system was established to comply with Specification 3.12. The initial failure to start when expected was due to corrosion on some of the engine start relay contacts. A, study has been undertaken to explore the possibilities of water proofing / shielding these components. The starter motor failure was caused by a loose connection at the' voltage regulator. The 1/2 B Fire Pump. fire. damaged components were replaced and the loose wire was retightened. The 1/2 B Fire Pump was tested and deemed operable on November 21, 198 (4) (Closed) LER 87028, Revision 00: Open Penetrations Found Through Fire Barrier Due to Personnel Error This item was discussed in Inspection Report 254/87027(DRP).

The licensee's corrective actions have been implemented and the resident inspectors will continue to monitor activities in this are (5) (0 pen) LER 87026, Revisions 00 and 01: Piping Support Outside Compliance with Safety Analysis Report Due to Construction Erro This item is being followed by a Region III inspecto (6) (Closed) LER 87029, Revision 00: Bus 13-1 Degraded Voltage 5 Minute Timer Found Out of Tolerance Due to Setpoint Drif On December 7,1987, while Unit I was in the refuel mode, it was. discovered that the five (5) minute time delay relay for the degraded voltage trip at Bus 13-1 was out of toleranc The degraded voltage five-minute time delay relay actuated at 319.5 seconds, 4.5 seconds outside of the 300 + 15 seconds

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allowed by Technical Specifications. Bus 13-1 provides emergency power to the 1A and IB Residual Heat Removal (RHR)

pumps and 1A core spray pump. Degraded voltage on Bus 13-1, for 5 minutes, will result in tripping the above pumps, removes l non-essential leads from the bus, and starts and loads the 1/2 '

Diesel Generator to Bus 13-1. There is one five minute time delay relay on this bus. The root cause of this incident was instrument setpoint drif The time delay relay was '

recalibrated and teste If the off normal voltage was severe enough to decrease to 3045 volts, additional undervoltage relays would have bypassed the degraded voltage timer and initiated starting of the 1/2 Diesel

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Generator, removed nonessential loads from Bus 13-1 and loaded the 1/2-Diesel Generator to Bus 13- (7) (0 pen) LER 87025, Revision 00: Control Room Habitability Design Basis Assumption Error Results in Exceeding Allowable Filter Efficienc This item was discussed in Inspection Report 254/87027(DRP).-

The licensee has hired consultants.to study and resolve this probita. This item is being followed by Region based inspector ~

(8) '(Closed) LER 87027, Revision 00: Control Room Yentilation Isolations Due to Toxic Gas Analyzer Trip This item was discussed in Inspection Report 254/87027(DRP).

The-licensee's corrective action appears to be adequate. The-residents will continue to monitor the Licensee's proposed *

action (9) (0 pen) LER 87033, Revision 00: Anticipated Transient Without Scram Instrument Sensing Lines Inadequately Supported Due to Cognitive Personnel Error and Inadequate Desig This item is being investigated by a region based inspecto (10) (0 pen) LER 87018, Revision 01: IB Recirculation Motor Generator Field Breaker Failure to Trip - Root Cause Undetermine This item is being followed by a region based inspecto Unit 2 (1) (0 pen) LER 87019, Revisions 00 and 01: Piping Supports Outside Compliance with Safety Analysis Report Due to Design Erro This item is being investigated by a region based inspecto i (2) (Closed) LER 87020, Revision 00: Unit 2 Reactor Scram Due to l Failure of Turbine Master Trip Solenoid Valv i This item was discussed in Inspection Report 265/87027(DRP).

l (3) (Closed) LER 88001, Revision 00: Reactor Scram Due to l Turbine / Generator Load Reject - Cause Undetermine This item was discussed in paragraph . Review of Routine and Special Reports (90713)

The inspectors reviewed the Monthly Performance Reports for the months of November and December 198 *

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outage planning. Additionally, plans are in place to increase the staffing in this department, which currently has four members, to a -

staff of eight personne '

12. Part 21 Followup (92703)

A Part 21 report was issued by General Electric.concerning certain types of HFA auxiliary relays. This station had purchased four HFA154 relays in 1983 to be used in a modification. The modification has not been completed nor will it be.in the near future. -The Service Action Letter (SAL) has been incorporated into the modification package, a reference to this SAL will be placed into the computer description of these relays, and each relay will be tagged to reference the SA .

This item will be tracked as an Open Item (254/87033-05(DRP)).

13. Open Items Open Items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part_of the NRC or licensee or both. One 0 pen Item disclosed during this inspection is discussed in Paragraph 1 . Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1)

throughout the inspection period and at the conclusion of the inspection on February 8,1988, and sumarized the scope and findings of the inspection activitie The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents / processes as proprietar ,

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No violations or deviations were identifie . Bulletin Followup (92703, 25026) IE Bulletin No. 85-03: Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Setting This bulletin requested that the Licensee develop and implement a program to ensure that switch settings on certain safety-related motor operated valves were selected, set and maintained correctly to accomodate the maximum differential pressure. expected on safet valves during both normal and off normal events within the design basis. The maximum time allowed for this activity was two years from the date of the bulletin, which was issued November 15, 198 As of February 6, 1988, the licensee has not finished all items as requested by this bulleti Failure to meet this commitment has resulted in a Notice of Deviation (254)87033-04(DRP) and 265/87033-03(DRP)). This bulletin will remain open pending completion of all requested items, IE Bulletin No. 87-02: Fastener Testing to Determine Conformance With Applicable Material Specification This bulletin was issued due to a concern over the potential use of inferior fasteners in nuclear power plants. The station was requested to review its procedures for receipt inspection of fasteners, obtain fastener samples and perform mechanical and chemical tests, report the results to NRC, and based upon the results describe any further actions necessary to assure that fasteners used in the plant meet the requisite specifications and requirements. The licensee provided the information requeste NRR will compile and evaluate the results of the tests. This bulletin will remain open pending NRR's evaluation . Administrative Controls Affecting Quality (42700)

Several drawings and procedures were checked for adequacy and accurac Errors found were brought to the attention of the licensee and are in the

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process of being corrected. No violations or deviations were identifie l 10. Radiation Control (71709)

Periodic inspections of plant radiological control conditions were made during the inspection period. Isolated instances of minor deficiencies were found and promptly corrected by plant personnel. No violations or deviations were identifie i 11. Outages (60710, 86700, 71711)

l The Outage Planning department assumed a larger role during this outage ,

than in the past refuel outage. The increased use of work planning !

programs during this outage is expected to provide a data base for future '

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