IR 05000440/1988004

Insp Rept 50-440/88-04 on 880223-0419.Violations Noted.Major Areas Inspected:Previous Insp Items,Operational Safety, Nonroutine Events,Maint,Surveillance,Esfs,Operational Safety Team Insp Findings & Allegations

ML20154Q421
Person / Time
Site:	Perry
Issue date:	05/27/1988
From:	Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20154Q411	List: IR 05000440/1988004 ML20154Q406 ML20154Q416
References
50-440-88-04, 50-440-88-4, NUDOCS 8806070015
Download: ML20154Q421 (19)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-440/88004(DRP)

Docket No. 50-440/88004 License 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: February 23 through April 19, 1988

Inspectors: X. A. Connaughton G. F. O'Dwyer Steven Ray Approved By:

Mc h R. Cooper, Chief Reactor Projects Section 3B

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g Date Inspection Summary Inspection in February 23 through April 19, 1988(Report No. 50-440/88004(DRP))

Areas Inspected: Routine unannounced inspection by resident inspectors of previous inspection items, operational safety, nonroutine events, maintenance, surveillance, engineered safety features, Operational Safety Team Inspection findings, allegations, onsite review comittee activities, physical security, and radiological control Plant status meetings between licensee and NRC regional management personnel were conducted on March 25, 1988 and April 15, 198 Results: Of the 11 areas inspected, one violation was identified in one area (failure to take required actions for inoperable APRM instrument channels -

Paragraph 4.b.); and one violation was identified in a second area (failure to measure valve stroke time with the required accuracy - Paragraph 8.).

An Operational Safety Team Inspection was conducted at the Perry site on March 14-25, 198 Initial inspector followup of OSTI inspection findings is documented in Paragraph 8. of this repor PDR ADOCK 0500o440 g DCD

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D_ETAILS Persons Contacted 2 Alvin Kaplan, Vice President, Nuclear Group I C.M.Shuster, Director,NuclearEngineeringDepartment(NED)

M. D. Lyster, General Manager, Perry Plant Operations Department (PP00)

• 1,2 R. A. Stratman, Manager, Operations Section, (PP0D)

1,2 V. K. Higaki, Manager, Outage Planning Section (PP00)

• 1 M. Cohen, Manager, Maintenance Section (PP00)

• 1,2 F. R. Stead, Director, Perry P' t Technical De W. R. Kanda, Manager, Technica. section (PPTD) partment (PPTD)

S. F. Kensicki, Technical Superintendent (PPTD) I L. L. Vanderhorst, Radiation Protection Section (PPTD)

• 1,2 E. M. Buzzelli, Manager, Licensing and Compliance Section (PPTD)

1 R. A. Newkirk, Manager, Technical Section (PPTD)

S. J. Wojton, Manager, Radiation Protection Section (PPTD)

2 E. Riley, Director, Nuclear Quality Assurance Department (NOD)

T. A. Boss, Supervisor, Quality Audit Unit (NQAD)

D. J. Takas, Manager, Mechanical Maintenance Quality Section (NQAD)

• Denotes those attending the exit meetir.g held on April 19, 198 Denotes those attending the March 25, 1988 plant status meetin Denotes those attending the April 15, 1988 plant status meetin . Licensee Action on Previous Inspection Findings (92701, 92702) (Closed) Violation (440/86020-01(DRS)): Inadequate test procedures caused offgas system charcoal adsorber fire (three examples). The first example involved Attachment 1 to the data sheet for Generic Procedure GEN-M-021. Attachment 1 (which delineated the testing)

did not take into account the factgthat the charcoal ignition temperature could be as low as 307 F. The inspector determined that appropriate licensee personnel had become cognizant of this fact (see also the closeout of unresolved item (4"0/86020-03(DRS))

in Paragraph 2.b of NRC Inspection Report 440/876.6). Also, administrative controls for the use of space heaters had been established (see clesecut of violation (440/86020-02(DRS)) in NRC Inspection Report 44e/87003).

The second example was that GEN-M-021 was an inadequate procedure for generating test instructions. This concern was resolved when the licersee cancelled GEN-M-021 on December 17, 1986. The inspector reviewed all other tests generated by GEN-M-021 and found them to be acceptable (see discussion and closecut of open item (440/86020-05(DRS)) in Paragraph 2.c of NRC Inspection Report 440/87016). The related concern that GEN-M-021-generated instructions did not require sufficient review was also resolved by its cancellatio . - - _ . - _ .

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The third example was that the Work Order (W0) which controlled the offgas vault refrigeration system testing activity was incorrectly designated safety class 5 (nonsafety-related) and, therefore, the W0 did not receive appropriate levels of review. Licensee personnel informed the inspector that they had reviewed all GEN-M-021-related W0s from November 1985 to August 1986 (16 cases) to verify that correct safety class designations were specified. Four additional W0s. associated with current work in progress were directly examined and an on-line computer terminal which provided direct access into the W0 data base was used by the inspector to verify safety classi-fications of additional current W0s. These reviews ensured that the safety class of the W0 was the same as the safety class of the item being worked as designated on the P rry Nuclear Power Plant (PNPP)

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QualificationList(Q-List).

The inspector had concerns about the licensee's process of designating W0 safety class by referring to only the PNPP Q-Lis Therefore, the inspector reviewed the W0s listed below to ensure that the safety class of each W0 was commensurate with the highest safety class of any item that could have been adversely impacted by the work. The inspector found all the below listed W0s to have appropriate safety class designation In response to the inspector's concern about the lack of a Nuclear Quality Assurance Department (NQAD) review for the WO involved with the offgas system charcoal adsorber fire, the licensee revised Perry Administrative Procedure (PAP)-0905, "Work Order Process" to require an NQAD review for any W0s which specifically address the remeval and/or disassembly of components which are: safety class 1, 2, 3, 4, SR, or MC, ASME Code related (Sections I, III, IV, VIII, and XI),

, Apprendix R fire barriers required for electrical separation, or for items which are to be welded onto safety related components.

' (Closed) Open Item (440/86020-04(DRS)): Concern that W0s may not '

j' receive proper safety class designations. Actions to close this item were documented in the closecut of example three of Violation (440/86020-01(DRS)) in Paragraph 2.a. of this inspection repor . (Closed) Open Item (440/86020-06(DRS)): Untimely notification of fire protection engineering personnel following offgas system l

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charcoal adsorber combustion event. The inspector reviewed Plant Administrative Procedure (PAP)-1911. "Fire Emergency," Revision 1, dated October 21, 1985. Temporary Change Notice (TCN)-004 to PAP-1911, dated February 2, 1987 provided clarification that, upon receipt of notification of a fire within the owner controlled area, Satellite Alarm Station (SAS) personnel shall notify designated members of the Fire protection and Safety Unit. This procedural enhancement provided a more explicit assignment of responsibility for the notificaticn of fire protection engineering personnel than was in place at the time of the offgas system charcoal adsorber fir Inspector interviews with personnel assigned to the SAS indicated that the notification requirements were clearly understood by those responsible for carrying them out. The inspector has no further concerns regarding this matte (Closed) Violation (440/87003-04(DRP)): Reactor Core Isolation Ccoling (RCIC) system inoperability not reported within the time requirements specified in 10 CFR 50.72 (b)(2). The inspector vecified by document review that corrective actions specified in the licensee's response letter, dated May 1, 1987, were implemented as follows: a Standing Instruction was issued on February 10, 1987, which clarified the reportability requirements for unplanned inoperability (failures) cf the Reactor Core Isolation Cooling (RCIC) and the High Pressure Core Spray (HPCS) systems, and; Plant Administrative Procedure (PAP)-0606, "Condition Reports and Irranediate Notifications," was revised to provide similar clarifica-tion regarding the designation of RCIC and HPCS as single train safety systems. The inspector noted that subsequent to implementa-tion of the forgoing corrective actions, the Reactor Core Isolation Cooling system had been reclassified as a non-engineered safety feature system. This reclassification was based upon a reanalysis of the control rod drop accident which did not take credit for RCIC system operation. The analysis was incorporated into the Perry Final Safety Analysis Report. Future unplanned RCIC system inoperability will therefore not te reportable pursuant to 10 CFR 50.72 and 10 CFR 50.73 as a single train safety system failur (Closed) Open Item (440/87003-06(DRS)): Administrative controls for motor-operated valve packing adjustments. This item was written to track future planned inspector reviews of this matter. This item was determined by NRC Regional office management to be included in future-planned inspection activities being tracked by NRC Inspection and Enforcement Bulleting (IEB) 440/85003-8B. This item, therefore, serves no purpose and is administratively close f. (Closed) Open Item (440/87003-07(DRS)): Adequacy of motor-operator sizing for Reactor Core Isolation Cooling (RCIC)/ Residual Heat Removal (RHR) steamline inboard containment isolation valve 1E51-F063. Subsequent to the identification of this item, the licensee implemented a design change which replaced the D.C. powered motor cperator for valve 1E51-fC63 with a larger A.C. powered operator. The de:ignation of valve 1E51-F063 as "normally closed"

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was changed to "normally open." These changes rendered the question concerning the sizing of the D.C. motor operator moot. Additionally, motor operated valve design, maintenance, and testing will be reviewed as part of future planned inspection activities related to IEB 440/65003-BB. Based upon the forgoing, this item is hereby administratively closed, (Closed) Violation (440/87004-02(DRP)): Containment entry through a personnel airlock with an inoperable door. The licensee's response letter dated May 20, 1987 reaffirmed that the root cause of this violation was miscommunication between ongoing and offgoing security personnel assigned to control access at the containment personnel airlock. In order to prevent recurrence, the licensee provided guidance to security personnel that verbal instructions were to be documented and thoroughly communicated during personnel turnover Additionally, verbal instructions were to be reported to the Security Shift Supervisor who, in turn, would confirm the accuracy of the instruction with the originator. The inspector has, in the course of routine inspection activities, observed implementation of this guidance by security personnel. Based upon these observations and a lack of repetitive occurrences since this violation was identified, the inspector has no further concerns regarding this matte h. (Closed) Violation (440/87012-03(DRP)): Inadequate performance of Main Steam Isolation Valve (MSIV) control logic preoperational testing. Corrective actions identified in the licensee's response letter dated October 8,1987 were reviewed and determined to be acceptable during a special Augmented Inspection Team (AIT)

Inspection conducted on June 17-20, 1987, and documented in NRC Inspection Report 440/87014(DRP). The AIT findings relative to this item were discussed in Paragraph 4. of the subject inspection repor i. (Closed) Violation (440/87012-12(DRP)): Failure to independently verify discharge path valve lineup prior to initiation of liquid radiological effluent release. Licensee corrective actions for this violation involved the counseling of individuals on the need to implement independent verification actions based upon equipment conditions at the time of liquid effluent releas Based upon a lack of repetitive occurrences subsequent to the time this violation was identified, these actions appear to have been effective and appropriate. The inspector has no further concerns regarding this l matte J. (Closed) Violation (440/87016-01(DRP)): Inadequate periodic test instruction results in reactor scra The inspector reviewed

Temporary Change Notice (TCN)-001 to Periodic Test Instruction i (PTI)-N2/-P0001, "Reactor Feedwater Pump Turbine Stop Valve Test,"

Revision 1, dated April 22, 1987. The subject TCN, which was I

implemented on October 2, 1987, required that the feedwater pump controller for the pump under test be placed in manual. This

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eliminated an increase in feedwater control system demand when a feedwater pump trip signal was generated during the test. Since the subject TCN was issued testing has been satisfactorily accomplished without perturbation of feedwater flow to the reactor vesse . Ojerational 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 equipirent 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 plar. The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control During a routine tour of containment during reactor operations, the inspector noted a number of isolation valves on branch lines (i.e. vent, drain, and test connections) associated with drywell penetrations which were closed but not locked closed. Additionally, the inspector identified one branch line isolation valve associated with a containment penetration, valve 1G41-F528, which was not locke Technical Specifications 3.6.1.1.1 and 3.6.2.1 required that containment and drywell integrity be maintained during Operational Conditions 1, 2, and 3. Associated surveillance requirements included verification once per 31 days that all containment and drywell penetrations not capable of being closed by operable automatic isolation valve and required to be closed during accident conditions were closed by valves, blind flanges, or deactivated automatic valves secured in position. Exceptions were provided for valves, blind flanges, and deactivated automatic valves which were located inside the containment, drywell, or the steam tunnel portion of the auxiliary building and which were locked, sealed, or otherwise secured in the closed position. These containment /drywell penetrations were required to be verified closed during each cold shutdown except that such verifications need nct have been performed more often than once per 92 day The inspector determined that valve 1G41-F528 and, apparently all drywell penetration branch line isolation valves were not required to be locked I or verified closed once per 31 days by licensee valve lineup and I surveillance test instructions. Instead, these valves were to be verified closed at the "each cold shutdown /92 day" surveillance frequency l which technical specifications specified for such valves that were

"locked, sealed, or otherwise secured in the closed position."

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The licensee's surveillance requirements were apparently based upon the presence of threaded pipe caps on the affected containment and drywell penetration branch lines. It was not clear to the inspector that the threaded pipe caps were acceptable for satisfying the technical specification provisions for locking, sealing, or otherwise securing such valves in the closed position. This matter will remain an unresolved item pending receipt of a technical specification interpre-tation from the NRC Office of Nuclear Reactor Regulation (440/88004-01(DRP)).

4. Followup of Nonroutine Events at Operating Pcwer Reactors (93702) Plant Shutdown Due to Increase in Drywell Unidentified Leakage On February 22, 1988 between midnight and 8:00 a.m., while operating at 98% reactor power, an increase in Drywell Unidentified Leakage from approximately 2.5 gallons per minute (gpm) to 4.6 gpm was detected by Drywell Floor Drain Sump instrumentation. The licensee reduced reactor power to approximately 12% and made a drywell entry to investigate the source of the increased leakage. The licensee detennined that the increased leakage was from Feedwater Maintenance Isolation Valve, IN27-F560 The valve had a bonnet-to-body leak and a collection shroud had previously been installed to capture the leakage and route it to the Drywell Equipment Drain Sump. The body-to-bonnet leak had increased and the collection shroud had failed to capture the increased leakag Even though unidentified leakage remained below the technical specification limit of 5 gpm, the licensee performed an orderly shutdown to repair the leak. Cold shutdown was achieved at approximately 11:10 p.m. the same day. The valve was repaired by injection of leak sealant by personnel from Team, Inc. into a collar over the body-to-bonnet seal area. This repair reduced the leakage to acceptable levels. The collection shroud was improved by installing another drain line from the enclosure to the Drywell Equipment Drain Sum The new drain line prevented water buildup inside the enclosure and allowed for potential increases in leak 69 The plant was restarted February 26, 198 Misadjustment of Average Power Range Monitor (APRM) Gain Settings During a plant startup on February 27, 1988, channel calibrations for the Average Power Range Monitoring system were performed in accordance with Surveillance Instruction (SVI)-C51-T0024, "APRM Channel Calibration Evaluation / Adjustment." Dased upon process

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computer-generated thermal power calculations, the APRM gain settings were adjusted such that the APRM thermal power readings were within 2% of calculated thermal power. The gain adjustments

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were completed at approximately 4:58 a.m. Following approval of surveillance test results, power escalation was continued. At 5:45 a.m., operating personnel identified a discrepancy between indicated l

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reactor power and plant electrical output (i.e. plant efficiency seemed unreasonably high).

Investigation disclosed that calculations of thermal power used for the APRM adjustments were in error due to incorrect values of average feedwater flow being utilized in the calculations. Hardcopy process computer printouts obtained over the timeframe in question identified the incorrect feedwater flow inputs on the "failed sensor list." Personnel performing the APRM gain adjustments were neither directed by procedure, nor apparently knowledgeable enough to check the process computer output for these indications of an invalid thermal power calculation prior to APRM gain adjustmen Following discovery of the invalid thermal power calculation, correct feedwater flow inputs to the process computer thermal power calculations were restored and thermal power was calculated to be 36%. Seven of the 8 APRM channels were reading excessively low (between 28% and 29.5%). The surveillance test was reperformed with satisfactory results on all APRM channels by 6:09 As a result of the APRM gain misadjustments, the APRM Neutron Flux High and Simulated Thermal Power High scram and rod block setpoints for APRM channels A through G were outside of their allowable values. RPS trip systems "A" and "B" had less than the required minimum number of operable channels for periods of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,14 minutes and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,12 minutes, respectively. Technical Specification 3.3.1 required, in part, that with less than the required minimum number of operable channels in both trip systems, the trip system with the least number of operable channels (in this case, the "A" trip system) be placed in the tripped condition within one hour and that the required minimum number of channels be restored to operable status or the plant placed in Startup within the following 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> The "A" trip system was not placed in the tripped condition during the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 12 minute time interval overwhich less than the required minimum number of channels were operabl Failure to take technical specification required actions with less than the required minimum number of APRM RPS trip channels operable is a violation (440/88004-02(DRP)).

5. Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components described 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 specifications. The following items were considered during this review:

the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable. Work requests were reviewed

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to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system i performanc I On April 18 and April 19, 1988, the following maintenance activities were  ;

observed / reviewed: determination of torque switch setting by M0 VATS and  ;

installation of correct torque limiter plate on the motor operator for  !

Standby Liquid Control (SLC) pump suction valve, IC41-F001A, authorized  !

by Work Order (W0) 86-15047, Revision 2. Initially, the mechanic '

performing the work improperly connected the M0 VATS equipment to the valve but recognized this when the oscilliscope appeared to indicate that the torque switch was tripping before the Thrust Measuring Device (TMD)

indicated any Bellville Washer Deflection. The mechanic failed to properly compensate for the differences between the valve's actual wiring and the sample wiring diagram (Figure 2) of General Electrical Instruction (GEI)-056, Revision 1, "Motor Operated Valve Analysis and Test System (M0 VATS) Testing." He consulted with a Senior Maintenance Technician, returned to the work area, properly connected the equipment and satisfactorily accomplished the testin The Senior Maintenance Technician indicated to the inspector that GEI-056 would be clarified by a Temporary Change Notice (TCN). The wiring diagram contained in gel-056 was typical for the actual wiring of approximately 85% of the valve The Senior Maintenance Technician, on his own initiative, was collecting other diagrams that were typical for the remaining 15% of the valves. He informed the inspector that he would be holding training sessions for all mechanics certified to perform M0 VATS testing to ensure that they can properly compensate for any wiring differences in the future. Until the TCN and the training are complete this will be tracked as open item (440/88004-03(DRP)). Monthly Surveillance Observation (61726)

On March 11, 1988, the inspector observed technical specifications required surveillance test SVI-E32-T5403-E, Revision 1, "Main Steam Isolation Valve (MSIV) Leakage Control System - Main Steamline B Pressure Functional for 1E32-N661E," and verified that testing was performed in accordance with 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

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directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne No violations or deviations were identifie . Engineered Safety Feature (ESF) Walkdown (71710)

During this inspection period, the inspector performed a detailed walkdown of train "A" and train "B" of the accessible portions of the Annulus Exhaust Gas Treatment (AEGT) Syste The system walkdown was

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conducted using Valve Lineup Instruction (VLI)-M15, Revision 4. Prior to conducting the walkdown, the inspector verified VLI-MIS against controlled Piping and Instrumentation (P& ids) for the AEGT Syste During the walkdown, both trains of the AEGT system were identified by the licensee as operable systems in accordance with technical specifications. The "A" train of the AEGT system was in operatio During the walkdown, the inspector directly observed equipment conditions to verify that housekeeping was adequate; no prohibited ignition sources or flammable materials were in the vicinity; valves and dampers in the system were installed correctly and did not exhibit gross packing leakage, bent stems, missing handwheels, or improper labeling; major system components were properly labeled, lubricated, and cooled and exhibited no leakage. The inspector verified that instrumentation was properly installed and functioning and that process parameter values were consistent with normal expected values; Valves and dampers were in their proper positions and local and remote indications were functional; essential support systems were operational; and the electrical and control board lineups were prope No violations or deviations were identifie . Operational Safety Team Inspection Finding - Motor Operated Valve Surveillance Testing (92701) Background On March 14-25, 1988, an NRC Operational Safety Team Inspection (OSTI) was conducted for Perry, Unit 1, in order to assess licensee performance in a number of areas and to determine whether or not the licensee had successfully made the transition from a construction /

preoperational test status to a ful?y operational status. This inspection was to be documented in NRC Inspection Report N /88200. Among the issues raised during the OSTI, one issue was identified as having potential immediate impact on plant eouipment operability. The issued involved potential valve limit switch setting inaccuracies resulting in non-conservative valve stroke time measurement During a review of valve stroke time data, one of the OSTI members noted that valve 1821-F067C exhibited a significant increase in measured stroke time between tests conducted in May and July 198 Further inquiry disclosed that, prior to the July 1987 stroke time measurement, the valve was subjected to Motor Operated Valve Analysis and Test System (M0 VATS) testing. A portion of the M0 VATS process involved valve limit switch adjustmen Based upon the significant change in limit switch setting and resultant change in measured stroke times for valve 1821-F067C, the OSTI member examined the licensee's methodology for verifying the accuracy of valve position indicatio The OSTI member concluded that, as outlined in the licensee's inservice test program

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implementing procedures, the licensee apparently only made static comparisons of locally observed valve full open or full closed status with remote full open or full closed position indication status. Such a methodology would not necessarily disclose premature valve position limit switch actuatio Inservice test results documentation similarly did not furnish objective evidence that the position indication tests were anything more than static comparison of valve status and position indication statu These findings led the OSTI member to question the accuracy of valve position indication in general and the accuracy of valve stroke time measurements which relied upon observation of remote valve position indication. Since the operability of certain valves was contingent upon meeting technical specification-specified valve stroke time limits, nonconservative valve stroke time measurements could have resulted in inoperable valves going undetected. Inspector followup

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of this issue was performed during this inspection to ascertain whether or not inoperable valves had gone undetected and to evaluate whether or not the accuracy of valve stroke time measurements was in accordance with the applicable requirements of 10 CFR 50.55a(g),

I technical specifications, and the licensee's inservice test program l for pumps and valves.

l Followup Inspection and Results

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The licensee assembled stroke time data for 21 motor operated valves which had not been subjected to M0 VATS testing. Comparison with technical specification stroke time limits disclosed that four valves had measured stroke times which were within one second of exceeding these limits. Definitive stroke time measurements for these valves were subsequently obtained by monitoring valve operator motor current during a valve actuation cycle. In each case the measured stroke time (motor run time) met technical specification limits and agreed with previous stroke time measurements (utilizing valve remote position indication) within two tenths af a secon This data indicated that these valves had close limit switch settings which reflected valve position with a high degree of accurac Inspector review of pre and post M0 VATS stroke time data for 64 valves which had been M0 VATS tested disclosed that, more often than not, valve stroke times obtained from remote position indication were shorter following M0 VATS testing. This trend was attributable to the fact that the close limit switches also functioned as open torqueswitch bypass limit switches. In order to ensure that the open torqueswitches were bypassed until the valves were fully unseated during an opening cycle, MOVATS testing systematically established close position limit switch settings that did not coincide with valve full closur _ _ _ _ _ - _ - _ _ _

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Based upon the motor run times obtained for valves which had been M0 VATS tested as well as the marginal valves which had not been M0 VATS tested, and the general trend in pre and post M0 VATS stroke time measurements, the inspector concluded the following: (1) the methodology for setting the close position limit switch prior to the introduction of M0 VATS testing closed position indication; (2) generally the pre-M0resulted in more precise-VATS inaccuracy of limit switch settings on valve 1821-F067C appeared to have been an isolated case, and not the result of a flawed limit switch setting methodology; (3) valve operability for all valve reviewed was not in question, and; (4) inaccuracies in close position limit switch settings resulting from M0 VATS testing were not accounted for in valve stroke time measurements and, in some cases, resulted in measurement inaccuracies in excess of those permitted by 10 CFR 50.55a(g).

The licensee's current approved inservice test program for pumps and valves which implemented the requirements of 10 CFR 50.55a(g) and more specifically, the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, 1983 Edition and Addenda through Summer, 1983, Subsection IWV, required that valve stroke times be measured to the nearest second, regardless of the valve's maximum allowable stroke time. The inspector noted that for valves with maximum allowable stroke times of greater than 10 seconds, the licensee's inservice test program stroke time measurement accuracy requirements were more restrictive than ASME Code requirements. The ASME Code required that stroke time measurements for such valves be accurate to the nearest 10% of the maximum allowable stroke tim The time between close limit switch actuation and valve full closure (based upon MOVATS data) resulted in stroke time measurement errors which exceeded either set of accuracy requirements for the following valves:

Required Measurement Max Allowable Accuracy (nearest Stroke Time 1 sec. or nearest 10%, Measurement Valve (Close Direction) as applicable) Error

, 1E12-F00738 15s 1 .75s -2.00s 1E12-F0024A 90s 1 4.50s -6.30s 1E12-F00248 90s i 4.50s -5.20s 1E22-F0012 Ss f .50s .54s 1E51-F0068 60s 1 3.00s -3.20s 1G33-F0001 15s 1 .75s .96s 1G50-F0272 20s 1 1.00s -1.80s l

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Failure to measure the forgoing valves' stroke times to the nearest 1 second or 10% of maximum allowable stroke time, as applicable, is contrary to 10 CFR 50.55a(g), applicable Edition and Addenda of the ASME Code,Section XI, Subsection IWV, the licensee's current, approved inservice test program for pumps and valves, and is a violation (440/88004-04(DRP)). _ Reviews of Allegations (99014)

(Closed) Allegation (RIII-87-A-119)

On August 31, 1987, the inspector received via the licensee's internal mail system, an anonymous letter which contained a number of concerns related to the conduct of a named individual in tne licensee's maintenance organization. The inspector determined by document review that the licensee's "Call for Quality" organization had earlier received an identical copy of the anonymous letter and was conducting an investigation to detennine the validity of the concerns and to effect corrective actions for concerns which were substantiated. Given the licensee's prior receipt of the anonymous letter and ongoing licensee investigative efforts, NRC Region III management determined that a NRC review of the results of the licensee's investigation would be conducted in lieu of wholly independent NRC followup investigatio Based upon the inspector's review of the licensee's investigation files concerning the anonymous letter and previous reviews of other Call for Quality files, the inspector determined that the licensee's handling of the issues raised in the letter was particularly thorough. The inspector noted that contributing to the thoroughness of the investigation was the fact that, during the investigation, three co-authors of the anonymously submitted letter identified themselves to Call for Quality investigators and availed themselves for followup interview Each of the 15 specific concerns contained in the letter were evaluated to determine whether or not they concerned quality / safety-related m?tter Those concerns which were determined to involve

. - quality / safety-related matters were investigated in detail to develop

! factual information in order to substantiate the concerns. In addition to licensee followup interviews with the co-authors of the letter, licensee investigative efforts included: interviews with the named individual targeted in the concerns; interviews with other individuals having first-hand knowledge concerning the occurrences upon which the concerns were based; reviews of applicable administrative controls governing activities covered in the concerns; review of licensee QA program corrective action documentation to determine if issues had been

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identified which were similar to those raised in the concerns, and; supplementary interviews of randomly selected maintenance personnel to determine if they were aware of any occurrences similar to those discussed in the concerns which were substantiated. The inspector verified that file information detailing the results of the forgoing investigative efforts supported the conclusions reached by the licensee for each concern. Each concern and associated licensee finding is summarized below.

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Concern 1 A Grade 1 and Grade 2 mechanic were instructed to top off the Division 3 batteries in Unit 2 prior to the repetitive task card being issue Finding This concern could not be substantiated in that the work was performed in accordance with the applicable surveillance test instruction and after the repetitive task card was issue Concern 2 A Grade 2 mechanic was instructed to train a Grade 1 mechanic to the M0 VATS and limitorque procedure Finding This concern could not be substantiate From a quality standpoint, both Grade 1 and 2 mechanics may perform maintenance to these procedures as long as they are properly trained and qualified. No procedure was violated in training additional personnel to perform this tas '

Concern 3 A Grade 1 mechanic was instructed to erect scaffoldin Finding Since erection of scaffolding is not a nuclear safety related activity, this concern is not considered quality related. However, our investigation revealed that mechanics do errect scaffolding from time to time when carpenters are not availabl Concern 4 The named individual instructed two Grade 1 mechanics to remove an operator from a valve prior to obtaining permission from the shift outage directo Finding While this concern was substantiated, Call for Quality found that the work was done in accordance with an approved work package and there were no procedure violations involving substandard work quality. The individual in question has been subsequently trained to the administra-tive procedure covering the work order process to ensure that future work is properly sequenced and performed with the cognizance of outage planning personne i

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.E Concern 5 A Grade 2 mechanic was instructed to enter a confined space with no safety man and no Confined Space Entry Permi Finding This concern could not be substantiated. Call for Quality found a Confined Spcce Entry Permit for this activity which was renewed each morning as work progresse Concern 6 The named individual wanted to send a crew into a confined space without a safety ma Finding This concern could not be substantiated. While the individual entertained the thought to change the area from a confined space, discussion with safety personnel revealed this was not possibl Consequently, the area remained classified as a confined spac Concern 7 A Grade 1 mechanic was instructed to transport oil to a diesel room without a Transient Combustible Permi Finding This concern could not be substantiated. A permit was issued to transport 55 gallons of oil to the Diesel Generator Building to cover this activit Concern 8 Two Grade 1 mechanics were instructed to perform two tasks which required an operator at the controlling motor control center bucket Finding l This concern could not be substantiated as a quality concern in that only l one job was completed at a time. Further, investigation revealed that the shift supervisor would not dispatch one man to cover two remote tasks since it would have been physically impossible to do s Concern 9 A Grade 1 mechanic was told to change a limiter plate using a mechanic assistant as a human ta .

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Finding This concern was substantiated. The individual was retrained to applicable procedures and was counseled by his superviso Concern 10 A Grade 1 mechanic was told to work on electrical equipment again using a mechanics assistant as a human ta Finding This concern was substantiated. The individual was retrained to applicable procedures and was counseled by his superviso Concern 11 A Grade 1 mechanic was told to make a design change on a sleeve 'sithout the use of an FC Finding This concern could not be substantiated. An FCR was written regarding the change and the work was completed in accordance with the instructions on the work orde Concern 12 A Grade 1 mechanic was instructed to obtain an Arc Chute Groundin Bar from Unit 2 without the use of a Material Transfer Authorization MTA).

Finding This concern could not be substantiated. Call for Quality found that the named individual had merely questioned whether an MTA was require A MTA was initiated but later voided and the necessary parts obtained from the warehouse. However, the individual in question was counseled on the importance of adhering to the MTA procedur Concern 13 A Grade 1 mechanic was instructed by the named individual to perform his work after waiting only 15 minutes for QC to cover a witness poin Finding While this concern could not be factually substantiated, the individual was, nevertheless, subsequently trained to procedures which required a 1/2 hour notification for QC inspectio '

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Concern 14 A Grade 1 mechanic was instructed to mill 20 thousandths from a site glass housing without an FCR. The mechanic refused to do s Finding While this concern was substantiated, Call for Quality found that an FCR was initiated and dispositioned by Engineering allowing the milling of the front plate. All work was accomplished in accordance with the work package and the FCR. The individual in question has been counseled regarding this concer Concern 15 A mechanic assistant was instructed to install three vacuum cleaners to the front of the Service Building using unistrut fastener Finding Since the Service Building is not a safety related structure this is not e

classified a quality concer Conclusion:

Based on the inspector's review, the only substantive violations by the named individual involved the use of "human tags" to isolate electrical equipment'for maintenance. These violations of licensee administrative controls concerned matters of non-radiological occupational health and safety practice and, as such, do not fall within the realm of NRC jurisdiction. The inspector did ascertain, however, that a formal grievance addressing this and other matters was filed by the mechanics pursuant to provisions of the existing labor agreement with the licensee on August 12, 1987. This matter had since been settle '

10. Onsite Review Committee (40700)

The inspectors reviewed the minutes of the Plant Operations Review Conunittee (PORC) meetings No.87-261 through 87-271,88-001 through 88-017,88-022 through 88-031,88-034 through 88-040 and 88-42 through 88-45 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 No violations or deviations were identifie . Physical Security Procedures For The Resident Inspector (71881)

During this inspection period, the inspectors observed / reviewed selected

licensee activities for conformance with the approved physical security plan. The inspectors reviewed security personnel staffing levels and i

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verified that individuals authorized by the physical security plan to direct security activities were provided for each shift. The inspectors observed that access control measures, including search equipment, protected area and vital area barriers, and security door locking devices were operational and in use. The inspectors observed that personnel and packages entering the protected area were properly searched in accordance with licensee procedures. The inspectors observed that persons granted access to the site were badged to indicate whether or not they had unescorted or escorted access authorizaion. Finally, by direct observation the inspectors determined that the effectiveness of detection assessment aids was maintained by the absence of obstructions in the isolation zone, adequate illumination of the protected area and protected area barrier, and operable video surveillance equipmen No violations or deviations were identifie . Radiological Protection Procedures For The Resident Inspector (71709)

Through discussions with licensee management, supervisory, and health physics personnel, and observation of licensee work planning activities, the inspectors determined that licensee personnel were aware of the ALARA program and that ALARA considerations were routinely considered in the planning of activities which involved occupational radiation exposur The inspectors also determined through monthly Plant Status Meetings such as the one described in Paragraph 14. of this report and review of the licensee's internally generated Monthly Performance Reports, that the status of meeting ALARA goals and objectives is periodically assessed and disseminated to affected plant personne No violations or deviations were identifie . Plant Status Meetings (30702)

On March 25, 1988, at the Perry site and on April 15, 1988 at the NRC Region III Office, NRC management met with CEI management to discuss the current status of the plant, recent events, and licensee initiatives to improve the quality of plant operating and maintenance activities. These meetings are being held on a periodic (initially monthly) basi . Open Inspection Items Open inspection items are matters which have been discussed with the l

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 Paragraph . Unresolved Items Unresolved items are matters about which more infonnation is required

! in order to ascertain whether it is an acceptable item, a violation or l a deviation. An unresolved item is identified in Paragraph 3.

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16. Exit Interviews (30703)

l The inspectors met with the '.icensee representatives denoted in Paragraph '

I throughout the inspection period and on April 19, 1988. The inspector summarized the scope and results of the inspection and discussed the l likely content of the inspection report. The licensee did not indicate i that any of the information disclosed during the inspection could be !

considered proprietary in natur l

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