IR 05000440/1987008

Insp Rept 50-440/87-08 on 870414-0601.No Violations or Deviations Noted.Major Areas Inspected:Previous Insp Items, Operational Safety,Nonroutine Events,Surveillance Testing, Maint,Startup Testing & Startup Test Results

ML20216F839
Person / Time
Site:	Perry
Issue date:	06/24/1987
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20216F791	List: IR 05000440/1987008 ML20216F787
References
50-440-87-08, 50-440-87-8, NUDOCS 8706300957
Download: ML20216F839 (12)

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U.S. NUCLEAR REGULATORY COMMISSION l

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REGION-III i Report No. 50-440/87008(DRP) l Docket No. 50-4401 License No. NPF-58 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101

' Facility Name: l Perry Nuclear Power Plant, Unit 1 Inspection At: Perry Site, Perry, OH Inspection Conducted: Apri1 ~ 14 through June 1,1987 Inspectors: K. A. Connaughton G. F. O'Dwyer Approved By: .R..C. Knop, Chief Y t(~ 2l&l Reactor Projects Section IB Date Inspection Summary

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' Inspection'in' April 14 through June 1, 1987 (Report No. 50-440/87008(DRP))

. Areas Inspected: Routine unannounced inspection by resident inspectors of previous inspection items, operational safety, nonroutine events, surveillance testing, maintenance, startup testing, and startup test.result "

Results

No violations or deviations were identified. During this inspection period, the licensee completed-Test Condition 2 startup testing and entered

' Test Condition 3 on May 15, 1987. On May 5, 1987, a meeting between'NRC and licensee management was held at the Perry site to discuss-plant status and

. licensee performance.

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DETAILS 1. Persons Contacted A. Kaplan, Vice President, Nuclear Operations Division C. M. Shuster, Manager, Nuclear Engineering Department (NED)

• M.'D. Lyster,. Manager, Perry Plar,t Operations Department (PP00)

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

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

• R. P. Jadgchew, General Supervising Engineer, Instrumentation and

' Controls Section (PP00)

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

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

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

• D. C. Jones, Operations Engineer, Licensing & Compliance Section (PPTD)

• G. A. Dunn, Compliance Engineer, Licensing & Compliance Section (PPTD)

• S. J. Wojton, General Supervising Engineer, Radiation Protection (PPTD)

• K. P. Donovan, Reactor Engineer, Technical Section (PPTD)

• E. Riley, Manager, Nuclear Quality Assurance Department (NQAD)

B. D. Walrath, General Supervising Engineer, Operational Quality Section (NQAD)

• V. K. Higaki, General Supervising Engineer, Maintenance & Modifications Quality Section (NQAD)

• Denotes those attending the exit meet'ng held on June 1, 198 . License,e Action"on Previous Inspection Findings (92701) (Closed) Unresolved Item (440/86023-02(DRP)): Unqualified wiring in Limitorque Motor Valve Operators. The NRC has deternined that no enforcement action will be taken for Environmental Qualification violations involving unqualified valve motor internal wiring. This i decision was based on the generic. nature of the deficiencies, j extenuating circumstances, and limited potential safety significance 1 of the violation (Closed) Open Item (440/86028-03(DRP)): Evaluation of scram discharge volume drain valve failur Licensee inspection of scram discharge volume drain valve IC11-F011 following a failure of the ,

valve to open on November 26, 1986, disclosed that the valve I coupling had come loose from the air operator. The coupling bolt was found loose, allowing the coupling halves to separate. The coupling was replaced and the valve was cycled twice to verify proper operation. As a result of this occurrence and scram discharge volume vent and drain valve coupling failures identified in General Electric Service Information Letter (SIL) 4.?2, the licensee retorqued all four scram discharge volume vent and drain l valve coupling bolts in accordance with the requirements of SIL 42 General Electric SIL 422 and NRC Information Notice 86-82 discussed

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scram discharge volume vent and drain valve failures resulting from l partial engagement of the manual valve actuators. In order t preclude such failures, the licensee revised Valve Lineup Instruction (VLI)-C11 to require that the manual actuator be placed and/or verified in the neutral position. Additionally, the lesson plans for operator training on the control rod drive hydraulic system were revised to provide a discussion of the valve failure mode involving partial engagement of the manual actuator and the method of verifying that the manual actuator is fully disengage The inspector verified that the foregoing actions had been accomplished by review of Condition Report 86-1035, Work Orders 86-15653, 86-15654, 86-15655, 86-15656, NRC/INPO Document Review Request for NRC Information Notice 86-82, and VLI-C11, Revision (Closed)OpenItem(440/86033-02(DRP)): Radiological controls for cutting, grinding, or welding contaminated or potentially contam-inated items. By review of Plant Administrative Procedure (PAP) i 0512, " Radiation Work Permits," including Temporary Change No. 1, dated May 22, 1987, the inspector determined that the licensee amended the list of conditions requiring a radiation work permit to include grinding, welding, or burning on contaminated or potentially contaminated systems or any evolution that may produce an airborn contamination hazard. The inspector has no further concerns in this are (Closed) Unresolved Item (440/87003-01(DRP)): Degraded motor leads in Limitorque DC Motor Operators. The NRC has determined that nu enforcement action will be taken for Environmental Qualification violations involving unqualified valve motor internal wiring. This decision was based on the generic natt're of tb deficien.ies, extenuating circumstances, and limited potential safety significance of the violation (Closed) Open Item (440/87003-02(DRP)): Voluntary entry into '

Technical Specification Limiting Conditions for Operation consistent with Technical Specification intent. The inspector reviewed Plant Administrative Procedure (PAP) 0205, " Operability of Plant Systems,"

Revision 4, dated May 15,.1987. The subject PAP was revised to include guidance concerning voluntary entry into Technical Specification Limiting Conditions for Operation. According to the PAP, multiple use of LCOs was not to be made for the purposes of completing maintenance or repair activities for equipment not governed by the applicable LCO. Multiple entries into Limiting Conditions for Operation for equipment governed by the applicable LC0 was not to be done except for unusual circumstances such as those involving personnel safety. The inspector has no further concerns regarding this matte . Operational Safety Verification (71707) General The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during i

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-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 components. 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 in accordance'with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control 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 procedures, Review of Reactor Level Instrumentation Performance During RCIC Injection As discussed in paragraph Se of Inspection Report 440/87003 and paragraph 10 ofLInspection Report 440/87004, the licensee observed level indication ' anomalies on= reactor water level instruments which utilized sensing'11nes connected to condensing chambers B21-D004A, B, C, and D during Reactor Core Isolation Cooling (RCIC) System-injection to the reactor vessel. To resolve this matter, the licensee nodified the~1nstrument unsing lines and performed retests with RCIC injectio The inspector reviewed various portions of the Emergency Response Information System (ERIS) plots for February 16, 1987 in which licensee personnel first detected level indication anomalie Discrepancies between expected levels and indicated levels varied from approximately 20 to 35 inches. The inspector also reviewed ERIS plots for Special Test, SXI-011 which also disclosed discrepancies of approximately 25 to 35 inche On April 27, 1987, the inspector witnessed the performance of various portions of Section 5 of Special Test, SXI-013, " Reactor Core Isolation Cooling (RCIC) Injection Following Level Instrument Reference Leg Nozzle Insert," Revision 0, which was performed to check for level instrumentation anomalies after modifications had been installed. All inspector test observation took place during the RCIC run at a reactor pressure of 250 psi. No level anomalies ,

were detected. Inspector analysis of the archived ERIS data plots )

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I, for level indication during the test confirmed that level indication was not adversely affected during RCIC injectio The inspector also reviewed ERIS plots for RCIC system quick starts with injection to the reactor vessel. RCIC pump flow to the reactor

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vessel increased from 0 gpm to at least 700 gpm in less than 22 seconds. The plot of wide range reactor water level instrument

output'showed an increase of only seven inches. This was expected-due. to the time delay associated with the feedwater Startup Level Controller and the sudden RCIC pump flow increas During; tests performe~d'on April 30, 1987, the RCIC pump was manually initiated and achieved 700 gpm flow within 30 seconds. On May 8,-

1987. the inspector reviewed ERIS plots of the outputs of reactor water level transmitters B21-N091A, B21-N091B, B21-N073C,'and i B21-N073D during these tests. The RCIC pump.ran at approximately

, 700 gpm for more than 1 5 minutes with a maximum increase in level indication of approximately 4 inches. Again, this was due to an expected actual reactor water level increase associated with-the addition of full RCIC flow. Within one minute, the Feedwater Control System restored reactor water level to that which existed prior to the RCIC. quick star Based upon the foregoing reviews, the inspector concluded-that modifications- to the impulse lines associated with reactor water level instrument reference legs were successful in correcting the

, level iristrument anomalies previously experienced during RCIC injectio Discrepancy Between Predicted and Actual Recirculation Flow / Core Flow Relationship On May 20, 1987, while operating at 50% power, channel checks were performed for the Average Power Range Monitor-(APRM) flow biased simulated thermal power high trip function. The channel checks were performed to verify measured core flow (total core flow) to be greater than or equal to " established core flow" at the recirculation loop flow used to establish the APRM flow biased trip setpoints. The channel check failed in that measured total core flow was less than " established core flow" for the existing >

recirculation loop flow. Power was reduced to approximately 30%

where the channel checks were again performed and satisfactory results obtaine At the time of this occurrence, " established core flow" was derived

. utilizing an analytically determined relationship between recircula-tion loop flow and total core flow. Startup testing had yet to be performed to determine the actual relationship between recirculation loop flow and total core flow over the entire ranges of recirculation and reactor power level. Upon coinpletion of required startup testing, the actual relationship between recirculation loop flow and total core flow will be evaluated for acceptability and utilized in lieu '

of the analytically derived relationship for establishing the APRM ;

flow biased simulated high thermal power setpoint !

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lAs an interim measure, the licensee reduced the APRM flow biased trip setpoint to account for the discrepancy between the predicted

.and measured recirculation loop flow / total core flow relationship .This matter is an'open item pending completion of-startup test dat acquisition and evaluation of results (440/87008-01(DRP)).

4. "Onsite Followup of Non-Routine Events at Operating power Reactors-(93702) General For each of the events discussed in Paragraphs 4b through 4d'below,

'the inspectors performed onsite followup inspection activities to gather factual information, to assess the events'for safety significance, and to evaluate diagnostic and. remedial actions taken by the licensee in response to each of the events. As applicable, these followup inspection activities included interviews with licensee personnel; review of operating, maintenance, and surveillance test records; pertinent plant data;-documented licensee event evaluations; and, associated corrective action documentatio .

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I May 1, 1987 Manual Scram Due to Decreasing Reactor Water Level On May 1, 1987, at approximately 5:27 p.m. while operating at 53%-

power, operations personnel received a No. 4 heater low level' alarm-due to decreasing water level in the hot-surge tank. Operators took manual control of the hot' surge tank level controller and attempted to increase flow to the hot surge tank'without success. An ..

auxiliary operator was dispatched to the hot surge tank. level-control valve to investigate. Reactor power was reduced to 38%

utilizing the reactor recirculation ilow control valves in ordar to reduce the rate of hot surge tank level decrease. At approximately 5:37 p.m., hot surge tank level reached the low-low setpoint causing the operating reactor feedwater booster pumps to trip. Per design, the' A and .B reactor feed pumps tripped and the motor-driven feedwater. pump autostarted. The reactor was manually scrammed immediately following the feedwater pump trips. Seconds later, hot surge tank level was recovered and the reactor feedwater booster pumps autostarted causing a waterhammer in the A and B feedwater pump suction piping. The motor-driven feedwater pump began feeding the reactor vessel and reactor water level was restored to within normal limits. Throughout the event, reactor water level remained above the low level 2 setpoint for high pressure core spray and reactor core isolation cooling system auto-initiatio Analysis and evaluation of the event conducted by the licensee, disclosed that hot surge tank level control valve IN21-F230 had failed at approximately 25% open due to the rupture of two associated control air lines. The control air lines were in contact with one another and had rubbed together.due to vibration, ultimately failing. The failed position of the valve resulted in insufficient flow to the hot surge tank. Auxiliary operators were unable to open the valve manually, because of misalignment between

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l the valve stem and-manual operator coupling. Licensee actions to repair valve IN21-F230 included replacenant of the control . air lines with armored flexible hose and realignment of the valve stem with the manual operator couplin The waterhammer experienced following the automatic starting of the reactor feedwater booster pumps damaged several feedwater pump suction piping hangers. -The licensee performed piping walkdowns and. inspections to determine'the extent of the damage and to identify items for-repair. 'As the result of these inspections, hangers 1N27-H130 and 1N27-H134 were identified as requiring rework prior. to operation of the A and B feedwater pump i The. inspectors directly observed that repairs to valve IN21-F230 and the identified pipe hangers were completed, as required, to support subsequent plant operation. The licensee initiated design changes to the feedwater booster pump control logic and the motor-driven feedwater pump trip logic to prevent future waterhammer events under similar circumstances. These design changes were scheduled for implementation during an outage scheduled 'for July 1987. As an interim measure, administrative controls were revised to provide

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instructions to plant operators on actions to be taken to prevent similar waterhammer occurrence c. May 24, 1987; Manual Scram Following Condensate Drain Line Failure On May 24, 1987, at approximately 3:05 a.m., the reactor was manually scrammed from 25% power following the identification of a condensate system leak which resulted from a weld failure on a 3/4 inch drain line at the upstream weldolet of a drain line isolation valve.. The drain line was located immediately upstream-of hot surge tank level control valve IN21-F230. Prior to the occurrence, feedwater control system startup testing was being performed which resulted in rapid cycling of valve IN21-F230 and higher than normal pipe vibratio Vibration of the drain line caused fatigue failure of the wel Following the scram, the leak was isolated and repair activities were initiated. The licensee removed the drain line isolation valve and installed a welded c,ap on the failed drain line. Another condensate drain line located immediately downstream of valve IN21-F230 was also cut and capped to preclude a similar failur Following these repairs, the plant resumed operation on May 25, 198 d. May 27, 1987, Reactor Scram Following Feedwater Controller Failure On May 27, 1987, at approximately 5:47 a.m. the reactor scrammed from 58% power on low reactor water level following a rapid reduction in feedwater flow to the reactor vessel. At the time of the occurrence, the motor-driven feedwater pump and the A l turbine-driven feedwater pumps were in service and being auto- 1 matica11y controlled by the master level controlle Following i the Scram, operators took manual control of the operating feedwater j l

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i pumps, maintained reactor water level above the low level 2 high 1 pressure core spray and reactor core isolation cooling systems- 'l actuation setpoint, and restored water level to within normal limits. Subsequent licensee investigation disclosed that the a feedwater flow decrease was attributable.to a failure of.the !

controller for the' motor-driven feedwater pump discharge valv ;

Controller output fai' led downscale causing the discharge valve to 1 go close Feedwater control system response was not rapid enough ]

to prevent the reactor Scram on low reactor water level. The failed I controller was replaced and the plant was returned to operation l later the same da q l

'No violations or deviations were identifie I Monthly Surveillance Observation (61726)

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On April 27, 1987, the inspector observed various portions of Surveillance Instruction (SVI)-G50-T5266, Revision 1, " Liquid Radwaste Release Permit," conducted for release permit No. 87-91. The inspector verified that the procedure was reviewed and approved in accordance l with licensee's administrative program. Appropriate approvals and j authorizations to initiate or continue the SVI were obtained before -l the SVI was commenced or continue ;

The multi-channel analyzer (MCA) utilized for the SVI did not detect any I pre-release activity in the sample. It was, therefore, not necessary

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to use the Meteorological Information Dose Assessment System (MIDAS) ;

computer to estimate maximum' doses possible to the public prior to or after the release. This was in accordance with Chemistry Instruction, CHI-54, Revision 0, " Radiological Effluent Data Reduction." The inspector reviewed documentation for liquid release Permit No.87-76L in which the MCA did detect pre-release activity in the sample. As ;

required, MIDAS was used to predict maximum possible doses to the public l prior to the release. MIDAS was then used to calculate these dose using more accurate data obtained during the actual release. The lab 4 coordinator explained that the actual exposures would always be less than I the~ predicted exposures because the predictive MIDAS calculations assume !

the following; maximum waste discharge flow rates, minimum dilution l factors, and minimum holdup tank decay times. The inspector observed that the numerical values obtained from both MIDAS calculations conformed to this explanatio '

On April 28, 1987, the inspector observed various portions of SVI-G50-T5266, conducted for release permit no. 87-92. The Radiological Waste Control Room (RWCR) technician terminated the discharge and flushed t and secured the system in accordance with the SVI. He also reviewed the test documentation and found no discrepancies. The inspector observed the system restoration, including independent verifications, from the ;

RWCR Panel and in the fiel l On May 19, 1987, the inspector witnessed various portions of the performance of Surveillance Instruction, SVI-E12-T2003, Revision 5,

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" Residual Heat Removal (RHR) C pump and Valve Operability Test." The I surveillance instruction was properly reviewed and approved by the licensee. Required administrative approvals were properly obtained prior to test initiation. All prerequisites and calibration verifi-cations were signed off in the instruction. The test was performed correctly by a sufficient number of qualified and knowledgeable personnel; well coordinated with adequate communications established among the test crew members. No Limiting Conditions of Operation (LCO)

were entered for this surveillance. Test data was accurate, complete, and the test results met the requirements of Technical Specifications ,

(TS). Surveillance test documentation was reviewed in accordance with '

administrative procedures. The inspector verified that the surveillance test was completed at the required frequency per TS requirement No violations or deviations were identifie . Monthly Maintenance Observation (62703)

On May 27 and 28, 1987, the inspector observed the rebuilding and bench testing of a valve motor operator, serial number (S/N) 333495, for use as a spare for a safety related valve. The procedure used was Corrective Maintenance Instruction (CMI)-0008, Revision 1, " Repair of'Limitorque !

Motorized Valve Operator Type SMB-0 through SMB-4 and SMB-4T," which had been approved in accordance with licensee's administrative progra Nuclear Quality Assurance Department (NQAD) Witness Points were established, where required, and observed. The Quality Control -(QC)

inspector was familiar with all aspects of the work and the maintenance activities were accomplished by personnel qualified in accordance with the licensee's administrative procedures. Replacement parts and

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materials being used were properly certified. Cleanliness requirements were in accordance with administrative procedure No violations or deviations were identifie . Startup Test Instruction (STI) Witnessing and Observation (72302)

l STI E51-014, Section 8.5, " Operation of RCIC From Remote Shutdown Panel" On May 6, 1987, the inspector observed various portions of STI E51-014, Section 8.5, " Operation of RCIC From Remote Shutdown i Panel," Revision 3, "RCIC." Before commencing the test, the Test '

Director ensured that the procedure being used by all test crew members was the latest revision.by ensuring that all Startup Test Change Notices (STCNs) recorded in the STCN log were entered into the procedure. The Test Director held a pre-test briefing and adequately stated test objectives, duties, precautions, and l limitations. Minimum crew requirements were met. All test '

conditions and initial conditions were signed off in the test procedure in accordance with procedural and Technical Specification (TS) requirement The test was conducted in accordance with TS 3.3.2. which allows RCIC indication and control to be transferred

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to the Remote Shutdown Panel' for up to one hour. Test equipment I required by the procedure was in service and calibrated to a common !

time base. . Test personnel actions and communications were correct j and timely during the test. A preliminary analysis was made to- -:

assure, proper plant' response during the test. The RCIC turbine did not isolate or t' rip during the. test. The level 1 acceptance l criteria for Section 8.5 of the STI were satisfied. All data collected by ERIS was archived and recorded for final analysis by licensee personne STI C51-012 Section 8.2, " Average Power Range Monitor (APRM)

Calibration at High Power"

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On May 7,1987, the inspector witnessed various portions of STI C51-012, Section 8.2, " Average Power Range Monitor (APRM)

Calibration at High Power". The inspector observed that all initial conditions were met. There were no Startup Test Change Notices (STCNs) in the STCN log that.were applicable to Section 8.2. Test crew staffing, communication, and coordination were adequate, correct, and timely. The STI required that the power level as indicated by the APRMs be adjusted as necessary to equal (+2%, -0%)

core thermal power level as determined by Computer Program, 00-3,

" Core Thermal Power and APRM Calibration." Four APRMs required gain adjustment and the' inspector concurred with the assessment by test personnel-that the level 1.and level 2 acceptance criteria were me Following the test, results were reviewed and initially approved by the test coordinator as required by the procedure. All data was collected as necessary for final analysis by licensee personne ' Technical Specification Limiting Conditions for Operation were met in that the min: mum number of operable APRMs per trip system was always' maintaine STI R43-031 Section 8.1, " Loss of Turbine Generator and Offsite Power" On May 10, 1987, the inspector witnessed the performance of STI R43-031, Section 8.1, " Loss of Turbine Generator and Offsite Power."

The inspector observed pretest briefings provided to test personnel as well as onshift operating personnel in which individual roles and responsibilities, lines of communication, expected plant response, and contingency actions were thoroughly discussed. The inspector verified that prerequisites and initial conditions had been established prior to test commencemen The test was initiated by tripping the main turbine generator and simultaneously tripping both startup supply breakers L10 and L20 while the reactor was operating at approximately 23% of rated core thermal power. The inspector independently verified the following expected plant responses; main steam line isolation valve closure, automatic starting and sequential loading of all three standby diesel generators, tripping of the main generator on reverse power, and the lifting and reclosure of safety relief valves. Actions by

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A operating and test personnel were observed to be well-coordinated, timely, and correc Preliminary evaluation of' test results indicated that all level 1 acceptance criteria were met and all but one level 2 acceptance criterion were met. The inspector noted that a Test Exception Report was initiated when two safety relief valves tailpipe temperatures did 'not return to within 10*F of their original temperature following reclosure as specified-in the' associated level 2 acceptance criterion. The test exception was slated to be evaluated by cognizant personnel, along with the rest of the test result No' violations or deviations were identified.

, Startup Test Results Evaluation (72301)

The inspector reviewed the completed and approved startup test results package for Startup Test Instruction (STI)-R43-031, Section 8.1, " Loss of Turbine Generator and Offsite Power," dated May 13, 1987. The inspector reviewed all test changes to verify that each was approved in accordance with licensee administrative procedures and that applicable portions of the procedure were anotated to identify those changes. The inspector

' determined that the test changes had not altered the basic objectives of the test as described in the Perry FSAR, Chapter 1 The inspector reviewed.the documented resolution for test exception report 208-1 which was initiated when two safety relief valve tailpipe temperatures did not return within 10 F of their original temperature after reclosure (this was a level 2 acceptance criterion). The inspector concurred with the licensee's determination that weeping of the safety relief valves did not constitute an operating problem, was acceptable for continued plant operations, and that no change to the FSAR or intent of .

the startup test was involved in the resolution. The inspector verified I the resolution had been accepted by appropriate licensee management, and that maintenance work orders had been initiated to repair the weeping safety relief valves during the first refueling outag The inspector verified that all data sheets were completed as required with individual test steps and data sheets properly initialed and date The test exception discussed earlier was anotated on the appropriate data sheet. The inspector verified that the test results package including the test summary and evaluation had been prepared, reviewed, and approved by appropriate licensee personnel and the NSSS supplier, General :

Electric, and that the evaluation specifically compared test results with established acceptance criteri No violations or deviations were identifie . Plant Status Management Meetings (30702)

On May 5, 1987, NRC management met with CEI management to discuss the current status of the plant and recent event The meeting was conducted

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at the Perry site. These meetings are being' held on a periodic )

'(initially monthly) basi The meeting included discussions of: the status of the plant; recent l Licensee Event Reports (LERs); corrective actions taken or planned to be'taken to preclude repetition; and, the schedule for future evaluation l 10. Open~ Inspection Items f Open inspection' 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. An Open inspection item disclosed during the inspection is discussed in 1 Paragraph 3 l 1 ExitInterviews(30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 throughout the inspection period and on June 1, 1987. The' inspector summarized the scope and results of the inspection and discussed the 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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