IR 05000440/1987004

Insp Rept 50-440/87-04 on 870304-0413.Violations Noted: Failure to Take Timely & Effective Corrective Action Re Diesel Generator Air Leaks & to Maintain Closed Outer Door When Inner Containment Personnel Airlock Door Inoperable

ML20213A128
Person / Time
Site:	Perry
Issue date:	04/20/1987
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20213A089	List: IR 05000440/1987004 ML20213A087 ML20213A107
References
50-440-87-04, 50-440-87-4, NUDOCS 8704270402
Download: ML20213A128 (12)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

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Report No. 50-440/87004(DRP)

, Docket No. 50-440 L; cense No. NPF-58 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name: Perry Nuclear Power Plant, Unit 1 Inspection At: Perry Site, Perry, OH Inspection Conducted: March 4 through April 13, 1987 Inspectors: K. A. Connaughton G. F. O'Dwyer Approved By: R. p, i

Reactor Projects Section IB Date l

Inspection Sunrnary Inspection on March 4, 1987 through April 13, 1987 (Report No.

! 50-440/87004(DRP))

Areas Inspected: Routine unannounced inspection by resident inspectors of

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previous inspection items,10 CFR Part 21 reports, operational safety, nonroutine events, engineered safety features, Licensee Event Reports, surveillance test activities, maintenanc.c activities, startup test activities, !

j and onsite review committee activities.

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Results: Of the 10 areas inspected, three violations were identified in one area (failure to take timely and effective corrective action with regard to diesel generator control air leaks, Paragraph Sb), (failure to comply with technical specification 3.6.1.3 requirement to maintain the outer door closed when the inner containment personnel airlock door had been declared inoper-able Paragraph Sc), and (failure to provide procedures appropriate to

. existing plant conditions, Paragraph Se). The first two violations were

considered for escalated enforcement action as Severity Level !!! violations;

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however, considering the surveillance test history prior to the diesel generator failures and the fact the requirements contained in the applicable TS LC0 were adhered to following the failures and considering the relatively l insignificant core fission product inventory and the brief period of time i involved in the containment integrity violation, both violations were i classified as Severity Level IV violations. A plant statue, meeting was

] conducted between licensee and NRC Region !!I management 1ersonnel on April 9, 1987. At the close of this inspection period, t1e facility was

conducting a reactor startup and entry into Startup Test Condition 2.

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DETAILS l

1. Persons Contacted A. Kaplan, Vice President, Nuclear Operations Division

• C, M. Shuster, Manager, Nuclear Engineering Department (NED)

M. D. Lyster, Manager, Perry Plant Operations Department (PPOD)

• D. J. Takas, General Supervisor, Maintenance Section (PP00)

• R. A. Stratman, General Supervising Engineer, Operations Section, (PPOD)  ;

• R. P. Jadgchew, General Supervising Engineer, Instrumentation and ControlsSection(PP00)  ;

F. R. Stead, Manager, Perry Plant Technical Department (PPTD)

• S. F. Kensicki, Technical Superintendent (PPTD)

• G. A. Dunn, Licensing and Compliance Section (PPT,1)

L. L. Vanderhorst, Radiation Protection Section (PPTD)

E. M. Buzzelli, General Supervising Engineer, Licensing and Compliance Section(PPTD)

• E. Riley, Manager, Nuclear Quality Assurance Depar*. ment NQAD)

• B.D.Walrath,)GeneralSupervisingEngineer, Section(NQAD Operationa

• Denotes those attending the exit meeting held on April 13, 198 l 2. Licensee Action on Previous Inspection Findings (92701)

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, (Closed)OpenItem(440/86023-06 Supervisory approval for exceedingovertimeguidelinespr(DRP)): i ior to overtime wark assignment TheinspectorreviewedPlantAdministrativeProcevure(PAP)-0110, t

" Shift Staffing," Revision 2, dated February 27, 1987. This  !

procedure was revised to require supervisory evalu.ition of employee hours of work prior to each overtime assignment. if, as a result of this review, it was determined that the employee's hours of work will exceed established overtime guidelines, the supervisor was required to document the matter via an Overtime Deviation Request form. The Overtime Deviation Regnest was required to be forwarded  :

to designated management personnel for documented approval. The '

completed Overtime Deviation Request form was considered a quality i record for the purposes of establishing compliance with licensee consnitments regarding plant staffing and hours of work for personnel !

performing safety related functions. The inspector lias no further i concerns in this are ' (Closed)UnresolvedItem(440/87003-05(DRP)): Divisions 1 and 2 diesel generators inoperable due to leaking valves in the control

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, air systems. Inspector review of the circumstances surrounding the i I

diesel generator failures was conoleted during this it,spection and '

l 1s discussed in Paragraph Sb of t11s report.

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l l CFRPart21ReportFollowup(92701)

(Closed) 10 CFR Part 21 Report (440/86006-PP)(DAR-270): By letter dated l June 30, 1986, the NRC was informed by BBC Brown Boveri, Inc. of a i potential defect involving electrically operated Type K600 and K800 l circuit breakers. Attachment A to the subject letter identified that

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these types of circuit breakers were utilized at the Perry Nuclear Power

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Plant. The potential defect involved improper routing of a control wire harness such that the harness was in contact with one of the racking gears. Contact with the helical gear could result in the wire harness l

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being cut, resulting in a failure of the charging motor, a loss of remote closing capability, a loss of shunt trip capability, or a loss of local and remote breaker open position indications. The inspector verified that the licensee was informed of the reported defect and had determined by inspection of all 18 affected circuit breakers for Unit 1 that the 10 CFR Part 21 Report was not applicable to Perry Unit 1. This

determination had been previously reported by the licensee in a letter dated August 20, 1986 from M. R. Edelman to J. G. Keppler. The inspector has no further concerns regarding this matte . Operational Safety Verification (71707)

The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during this inspection period. The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of Limiting Conditions for Operation associated with affected component Tours of the intermediate, auxiliary, reactor, and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of maintenance. The inspectors by observation and direct interview verified that the physical security plan was being implemented j in accordance with the station security pla [

The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control .

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These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedure violations or deviations were identified.

l Onsite followup of flon-Routine Events at Operating Power Reactors (93702) _ General for each of the events discussed in Paragraphs Sb through f below, the inspectors perfonned onsite followu) inspection activities to

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gather factual information, to assess t1e events for safety

) significance, and to evaluate diagnostic and ren.edial actions taken by the Itcensee in response to each of the events. As applicable,

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these followup inspection activities included interviews with Itcensee personnel; review of operating, maintenance, and surveillance test records; pertinent plant data; documented licensee event evaluations; and, associated corrective action documentation, b. Division 1 and 2 Diesel Generator Inoperability As discussed in Paragraph 5d of NRC Inspection Report 440/87003,on

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February 27, 1987 the Divisions 1 and 2 standby diesel generators

were declared inoperable following surveillance test failures due to l 1eaking valves in the diesel generator control air systems. During l this inspection period, the inspector performed additional reviews to determine whether or not the diesel generator failures were attributabic to inadequate or untimely corrective actions by the

[ licensee. At the inspector's request, the licensee provided a chronology of diesel generator corrective maintenance and routine surveillance test activities for the time period beginning on November 29, 1986 through February 27, 198 On December 4,1986 a work request was written which identified that Division 2 diesel generator control air valve 1R43-F115B exhibited leakage. Subsequently, on December 30, 1986 and February 1,1987, the Division 2 diesel was demonstrated operable by routine

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surveillance testing. During these tests, licensee personnel did not observe low control air pressure alarms which would have been indicative of further control air system degradation. However, the presence or absence of low control air pressure alarms during these tests could not be established by document review, since surveillant.e test procedures did not require observation and documentation of all alarms actuated during the test On February 10, 1987, a work request was written which identified that Division 1 diesel generator control air valve 1R43-F110A exhibited leakage. On february 11, 1987, the Division 1 diesel was demonstrated operable by routine surveillance testing. Again, the presence or absence of a low control air pressure alarm during this test could not be established by ducument review. Licensee personnel involved in the test did not observe any such alar The work requests which identified control air system Icalage on the Division 1 and Division 2 diesel generators were prioritized in accordance with licensee administrative controls as Priority 3 work requests. Per the licensce's administrative controls Priority 3 work requests were characterfred as follows:

" Repair within one week. Delay can be accepted, however, the listed plant equipment may fail or deteriorate with continued operation, and may ultimately result in impairment of operation."

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I Based upon the foregoing information, the inspector concluded that

, the delay in processing the work requests was based upon licensee f

judgement that the control air leaks were not severe enough to impair diesel operability as demonstrated by surveillance testing

! prior to the failure . In light of the importance of the diesel generators' safety

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function, the existence of control air leaks on both Division 1 and Division 2 diesel generators, and the lack of objective evidence concerning control air system performance trends, the delay in processing the work requests represented a failure to take timely and effective corrective action and resulted in the February 27, 1987, diesel generator failures. Failure to take timely and effective corrective action is contrary to 10 CFR 50 Appendix D, Criterion XVI, and the licensee's Quality Assurance Plan,Section16,andisaviolation(440/87004-01(DRP)).

c. _ Breach of Containment Integrity While in Operational Condition 1 On March 14, 1987 between 8:00 and 8:15 p.m., a non-licensed operator informed shift supervision of a leaking pressure equalizing valve on the upper containment airlock. Following this report, a security guard stationed at the upper personnel airlock was instructed by control room operators to prohibit airlock use. At I approximately 8:45 p.m., the security guard was relieved by another guard. The oncoming security guard was not informed by the offgoing guard of prohibition on airloct use. Five minutes later, at 8:50 :

j p.m., the inner door of the upper airlock was formally declared

] inoperable due to the equalizing valvo leakag At 9:00 p.m., a chemistry technician entered containment through l the upper personnel airlock, unchallenged by the security guard on duty. The containment entry involved opening and closing the outer ,

personnel airlock door. At 9:21 p.m., the same chemistry technician i exited containment through the upper airlock and again, cycled the outer door, open and close At approximately 9:30 p.m., control room operators learned of the chemistry technician's unauthorized containment entry through the upper personnel airlock and documented the occurrence via Condition Report 87-139. By 9:50 p.m. the outer door of the upper personnel afriock was locked closed in accordance with Technical '

Specification On March 17, 1987, following the performance of local Icak rate testing and evaluations of results, the licensee confirmed the unacceptability of the equalizing valve leakage and detemined that the unauthorized containment entry on March 14, 1987, resulted in

momentary losses of containment integrity and were reportable in

! accurdance with 10 CFR 50.72. The llconsee made the required NRC

! notification within the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following this determinatio l

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Technical Specification 3.6.1.3 required, that, with one primary containment airlock door inoperable, maintain at Icast the o>erable airlock door closed and either restore the inoperable airloc( door

, to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the operable airlock door closed. Entry and exit through the upper personnel airlock on March 14, 1987, after the inner door had been declared inoperable was contrary to this requirement and is a violation (440/87004-02(DRP)). I

Temporary Loss of Control Room Annunciators and Declaration of An '

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i At approximately 9:14 p.m. on March 16, 1987, while in Operational Condition 1 with reactor power at 297., non-safety related DC ,

electrical bus DIA was inadvertantly deenergized when a non-licensed operator restoring reserve charger supply breaker DIA08 to service, dropped a racking tool which struck adjacent DC bus supply breaker D1A03, causing it to trip. This resulted in a loss of all control room annunciators, trips of various pieces of operating equipment,

. and the loss of a number of balance of plant instrument channel i llouse electrical loads auto transferred from the auxiliary I

transformer to the startup transformer, reactor recirculation purrp l "A" tripped, and reactor feedwater pump "A" (the only operating

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l feedwater pump) began feeding the reactor vessel at an increas< ng i j rat ! Operators took manual control of the feedwater system and stabilized reactor IcVel within operating limits. The non-licensed operator

, imediately contacted the control rocin, informed licensed operators i of what had happened and was instructed to reclose breaker 01A0 Bus DIA was thus restored at approximately 9:19 p.m. Following restoration of bus 01A, operators restarted reactor recirculation .

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pump "A" and returned affected equipment to servic In accordance with the licensco's Emergency Plan, an Alert was

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declared at 9:20 p.m. do to the temporary loss of control room annunciators. All notifications were made in accordance with the i

licensee's Emergency plan and the Alert was terminated at 9:54 p.m.

' Subsequent licensee investigation of plant response to the event determined that equipment had performed per desig Reactor Water Level Decrease and Manual Scram

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At approximately 9:52 a.m. on March 30, 1987 while in cold shutdown, operating personnel observed reactor water level decreasing at a

rate of approximately 1 inch per minute. Dased u system valve lineups and activities in progress, pon a determined it was review of that water was draining from the reactor vessel via the "B" reactor  ;

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watercleanup(RWCU)systemreturnlineintothe"B"feedwaterline '

l which had been previously drained for flange leak repairs. The  !

level decrease was initiated when non-11consed operators unisolated

, the "B" reactor water cicanup roturn line as specified in the )

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i restoration section of togout No. 1 87-1326 associated with the ,

i feedwater system outage!. l

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At approximately 10:05 a.m., licensed operators began isolating the RWCU system using remote manual isolation valves. At approximately 10:00 a.m., with reactor water level at 180 inches, a manual scram was inserted in anticipation of an automatic scram signal at a watt.r level of approximately 178 inches. The level decrease was terminated at a water level of approximately 179 inches. With the RWCU system isolated, water level was restored to normal limits (approximately 197 inches) in the following 12 minutes. The event was attributed to a procedural deficiency in that the tagout restoration checklist should have left the "B" RWCU return line isolate Inspector review of this event included a review of a similar event which occurred on March 25, 1987. At approximately 5:22 a.m. on that day, during initiation of shutdown cooling utilizing the "B" residual heat removal subsystem, a level decreasu of approximately 68 inches was observed. Reactor water level stabilized at IPO inches which was above the low level 2 scram setpoint of approximately 178 inches. Licensee investigation detemined that water had drained from the reactor via the shutdown cooling return line into the "B" feedwater line and the long cycle cleanup return lin At the time of both of these events, valve 1B21-F0658, which would nomally have been closed to isolate the remainder of the feedwater system from the "B" RWCU return line and the "B" residual heat removal shutdown cooling return line, was open. The volve was earlier placed and maintained in the open position in accordance with the disposition of a nonconformance report which had been written to document observed damage to the valve stem. The nonconformnce evaluation concluded that if required to stroke, the valve would perform acceptably. Immediate repair was, therefore, not required. However, the evaluation further stated that if the valve were stroked, repair would be required to assure continued valve operabilit The off-normal valve position and the potential for draining the reactor vessel via the "B" shutdown cooling subsystem and the "B" reactor water cleanup system return lines had not been acknowledged in applicable operating instructions in use during the March 25, 1987 event. While the off-normal valve position may have been recognized in the preparation of the restoration checklist in use at the time of the March 30, 1987 ovent, the fact that the unisolated portions of the feedwater system were drained was apparently not considered. The inspector noted that the off-normal position of valve 1021-F065B contributed in wholu or in part to both events, failures to acknowledge existing plant conditons in the procedures which caused these events rendered the procedures inadequat :

This is contrary to 10 CFR 50, Appendix 0. Criterion Y and the l licensee's Quality Assurance Plan, Section 5, and is a violation !

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. Division 2 Diesel Generator (DG) Inoperability At approximately 5:00 p.m. on April 2,1987, while in Operational Condition 4, a workman damaged three cables for Division 2 DG auxiliary systems while drilling a hole through a penetration seal in the DG room wall to facilitate installation of a new cabl The licensee subsequently determined that the cables were for the DG building ventilation fans and an emergency service water flow indicator. The licensee declared the Division 2 DG inoperable at 6:10 p.m. The Division 1 DG was also inoperable at the time of this occurrence due to ongoing maintenance activitie The inspector verified by review of Itcensee actions, that the licensee complied with applicable Technical Specification provisions. The Division 1 DG was returned to operability at approximately 1:48 a.m., April 3,1987. The licensee's investi-gation into the root cause and corrective actions for this event, including measures to prevent recurrence, is ongoing. This matterwillbetrackedasopenitem(440/87004-04(DRP)).

6. EngineeredSafetyFeature(ESF)SystemWalkdown(71710)

During this inspection period, the inspector performed a detailed walkdown of the train "A" and conrion components of the Emergency ClosedCooling(ECC) System. The system walkdown was conducted utilizing Valve Lineup Instruction (VLI)-P42. Revision 4. dated January 27, 1986. Prior to conducting the walkoown, the inspector verified VLI-P42 against controlled Piping and Instrun.cntation Diagrams (PalDs) for the ECC system. No discrepancies were identified as a result of this verifications however, during the system walkdown, the inspector discovered that a valve labeled as the ECC Punp A Suction Line Drain Valve IP42-f650, was not in the location s pecified by Pt.10 D 302-621 Revision L. The inspector notified tio system engineer who, in turn, wrote a field Chango Request (FCR) to have licensee engineering personnel resolve the discrepancy. This will be tracked asopenitem(44L/87004-05(DRp)).

During the system walkdown, the inspector directly observed equipment conditions to verify that hangers and supports were made up properly; appropriate IcVels of cleanliness were buing maintained; piping insulation, heaters, and air circulation systems were installed and operationals with the exception noted above, valves in the system were installed in accordance with applicable PalDs and did not exhibit gross packing Icakage, bent stems, missing handwheols, or improper labelings and, that major system components were properly labeled and exhibited no leakage. The inspector verified that instrurentation associated with the system was properly installed, functioning, and that significant process parameter values were consistent with normal expected values. By direct visual observation or observation of remote position indication, the inspector verified that valves in the system flow path were in the correct positions as required by VLl P42; that where required, power was available to the valves valves required to be locked in position were locked; and, that pipe caps and blank flanges were installed as require ._ -_ _ _

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l At the time of the system walkdown, the ECC system was being maintained l in a secured status in accordance with system operating procedures. The inspector verified that documentation required )y licensee administrative procedures pertaining to temporary alterations and equipment tagouts for maintenance was consistent with the system's declared statu No violations or deviations were identifie . Licensee Event Reports Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, imediato corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specification LER 86060-0 Incorrect Classification of D/G Start Results in Missed D/G Surveillances LER 86063-0 Disconnection of Division 3 Dattery results in Technical Specification Violation LER 86065-0 Misunderstanding of Technical Specification Results in Missed RACS Action Requirement LER 86071-0 Transfer of RPS Bus Power Supply Results in Unexpected 00P isolations LER 86075-0 Alternate Sampling Equipment Deenergized Resulting in Technical Specification Violation LER 86078-0 Maintenance Technician inadvertantly Decnergites Dus Causing CSF Actuation LER 86081-0 Suryc111ance Instruction Deficiency Results in MSL

! solation LER 86084-0 Missing Rounds Log Page Results in Violation of Instrumentation Technical Specification LER 86087-0 Inadequate Operating Instruction Results in Technical Specification Violation LER 86090-0 Failure to Review Electrical Diagrams Results in RCIC System Isolation LER 86091-0 Inadvertent Dreaker Actuation Results in 00P ! solation The events described in LERs 86060-0. 86063-0. 86065-0. 86075 0 86084 0, and8600/0involvedviolationsofTechnicalSpecificationregulrement The inspuctor reviewed these occurrences for significance, method of identification, timeliness and adequacy of licensee corrective actions, and to determine whnther nr not the violations were repetitive of previous violations. The inspector's review concluded that theso

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violations were identified by the licensee in the course of the performance of activities mandated by existing administrative controls, the violations were of minimal safety significance, corrective actions were prompt and appropriate based upon identified root causes, and the circumstances and root causes contributing to the violation were not repetitive of those associated with previous violations. The inspector will continue to monitor Itcensee performance in these areas and evaluate future identified violations in light of these violations and licensee actions taken to prevent recurrenc D. MonthlySurveillanceObservation(61726)

On March 31, 1987, the inspector observed various portions of Surveillance Instruction (SVI)-B21-T0189-R, "ECCS/HPCS Drywell Pressure High Channel Functional for ID21-N667R," Revision The SV! had proper licensee review and approval signatures. The InstrumentationandControl(!&C)Technicianobtaineddocumented authorization from the unit supervisor to begin prerequisites and the supervising nperator's aut1orization to commence the test, as required. The test instrumentation used was within its specified calibration interval. The inspector observed that the channel was properly removed from and restored to service and that all LCOs were met. Test data was properly recorded, reviewed by the IAC personnel supervisor, satisfied Technical Specification requirements, and met the SV! schedul No violations or deviations were identifie . MonthlyMaintenanceObservation(62703)

On March 26, 1987, the inspector observed various work activities authorized by Work Order (WO) 87-2257 which included inssection of the hydraulic lif ters in the rocker assemblics for some of t1e valves on the Division 1DieselGenerator(DG). Valve No. 6 had previously exhibited a higher than normal noise level, which was indicative of possible lifter problems. Two hydraulic lifters associated with valve No. G were found to have hairline cracks and were replaced. The hydraulic lifters of othur valves were also inspected and replaced as necessar The inspector found the tagouts and authorizations to commence work satisfactory. The work was performed adequately in accordance with ProventiveMaintenanceInstruction(PMI)-019." Division 1and2EDG Rocker Arm and Valve Lif ter Maintenance,", Revision 2, which had received required reviews and approvals. Work was performed by a qualified Grado 1 Mechanic also skilled as a Machinist, as required by licensee administrative procedures. Radiological Controls were unnecessary and fire protection measures were sufficient. Replacement parts were adequately controlled. Quality control witness points were established and observe No violations or deviations were identifie '

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O 10. Startup Test Instruction (STI) Witnessing and Observation (72302)

During the inspection period, the inspector witnessed various portions of the following tests, including establishment of prerequisites, and preliminary post-performance analyses:

STI-C51-012 "APRM Calibration " Revision 2, Section 8.2, "APRM Calibration at High Power."

STI-C91-013 " Process Computer " Revision 2. Section 8.3, "LPRM Calibration and TIP Plotter Adjustment." Section 8.9,

"DynamicSystemTestCase(DSTC)-LPRMCalibration."

STI-C91-019 " Core Performance Test," Revision 3 Section 8.1, " Core Thermal Power Determination."

STI-821-026 " Safety Relief Valves (SRV)," Revision 1, Section "SRV Flow Test."

STI-C61-028 " Shutdown from Outside the Control Room," Revision 2 Section 8.1, " Shutdown from Outside the Control Room,"

and Section 8.2, "Cooldown from Outside the Control Room."

The inspector observed that test procedures of the latest revision were available and in use by all appro)riate crew members. Test crews were adequately staffed, sufficiently (nowledgeable, and their actions were properly coordinated. All test prerequisites and initial conditions were met or waived in accordance with test program requirements. Permanent plant equipment and test equipment required by the procedures were in service and where necessary, were calibrated to a common time bas Technical Specification LCOs appeared to be adhered to at all time When an SRV tail pipe pressure switch did rot actuate as required upon its SRV being opened during STI-021-026, it was declared inoperable and the appropriate LCO was entered at 7:35 p.m. on March 15, 1987. Licensee investigation revealed that the pressure switch had been isolate Operating personnel placed the switch back in service, restoring operability, and the LCO was exited. Acceptance Criteria for this test section were met or discrepancies were documented for processing in accordance with the licensee's startup test progra For STI-021-026, Section 0.2, the inspector independently reviewed the EmergencyResponseIndicationSystem(ERIS)MainGeneratorGrossMWe '

grapis which recorded positive indication of steam flow through each actuated SRV. From this review, the inspector concluded that the Level 1 !

Analysis acceptance criteria of the test were satisfied. The inspector concurred with the licensee's preliminary test results review that the Level 2 acceptance criteria were also satisfie I The inspector reviewed and concurred with the licensee's preliminary evaluation that the level 1 and level 2 acceptance criteria were satisfied with no exceptions for Section 8.1 of STI-C91-01 ,

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l The licensee noted while reviewing the test results of STI-C61-028 that level indication anomalies on channels B and C of the Wide Range Reactor Water Level instrumentation occurred during RCIC operation similar to

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those noted previously on channels A and D. The B and C channels will l therefore be subject to ongoing Ifcensee investigations and corrective actions to explain and/or eliminate the RCIC/ reactor water level instrument interaction phenomeno No violations or deviations were identifie . Onsite Review Connittee (40700)

The inspectors reviewed the minutes of the Plant Operations Review Committee (PORC) meetings No.87-028 thru 87-044, conducted prior to and during the inspection period to verify conformance with PNPP procedures and regulatory requirements. These observations and examinations included PORC membership, quorum at PORC meetings, and PORC activitie . Plant Status Management Meetings (30702)

On April 9,1987 NRC management met with CEI management to discuss the ;

current status of the plant and recent events. The meeting was conducted (

at the Perry site. These meetings are being held on a periodic (initiallymonthly) basis,  ;

The meeting included discussions of: the status of the plant; recent i l

Licensee Event Reports (LERs); corrective actions taken or planned to be taken to preclude repetition; and, the schedule for future evaluations. The meeting also included a discussion of NRC enforcement considerations pertaining to the violations discussed in Paragraphs 5b i and S . Open Inspection Items

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Open inspection items are matters which have been discussed with the r licensee, which will be reviewed further by the inspector, and which ;

involve some action on the part of the NRC or licensee or both. Open <

inspection items disclosed during the inspection are discussed in Paragraphs 5f and . ExitInterviews(30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 throughout the inspection period and on April 13, 1987. The inssector sunenarized the scope and results of the inspection and discussed tie likely content of the inspection report. The licensee did not indicate -

that any of the information disclosed during the inspection could be considered proprietary in natur !?

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