IR 05000440/1997013

Insp Rept 50-440/97-13 on 970826-28.No Violations Noted. Major Areas Inspected:Followup to Maintenance Rule Baseline Insp

ML20217E757
Person / Time
Site:	Perry
Issue date:	09/29/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20217E753	List: IR 05000440/1997013 IR 05000442/1997013
References
50-440-97-13, NUDOCS 9710070177
Download: ML20217E757 (8)

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U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket No.: 50-440 License No.: NPF 58 Report No.: 50-440/97013(DRS) .

I Licensee: Cleveland Electric illuminating Company

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Facility: Perry Nuclear Power Plant Location: 10 Center Road Perry, OH 44081 Dates: August 26 28,1997 Inspector: Martin J. Farber, Reactor Inspector, Rlli f

Approved by: James A. Gavula, Chief Engineering Speciaibts Branch 1 -

Division of Reactor Safeiy 9710070177 970929 PDR ADOCK 05000440 1 PDR

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EXECUTIVE SUMMARY Perry Nuclear Power Plant, Unit i NRC Inspection Report 50-440/97013(DRS)

This inspection was a followup to the Maintenance Rule Baseline Inspection conducted November 4 - 8,1996, which examined the licensee's implementation of 10 CFR 50.65,

" Requirements for Moritoring the Effectiveness of Maintenance at Nuclear Power Plants." The repast covers a three-day on site inspection by one regional inspecto Maintenance

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Five findings involving performance criteria, on-line risk assessments, structure monitoring, and corrective action were closed. One finding, related to risk significance determinations, remained ope *

The licensee's corrective actions were acceptable and, with the exception of one calculation error, were properly implemente .

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Report Detallt introduction This inspection was a followup to the Maintenance Rule Baseline inspection conducted November 4 - 8,1996, which examined the licensee's implementation of 10 CFR 50.65,

" Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The report covers a three-day on site inspection by one regional inspector, ll. Maintenanna M1 Conduct of Maintenance (62706)

The primary focus of the inspection was to review the licensee's corrective actions for the issues identified during the Maintenance Rule Baseline Inspection, which was conducted November 4 8,1996. This insp6ction was performed by one regional inspecto M8 Miscellaneous Maintenance issues (92902)

M8.1 {Q0en) Unresolved item 96014-02(DRS) Potenial Misclassification of Risk Significance Calculation 6.16, " Determination of Level 1 PSA Safety Significant SSCs for the PNPP Maintenance Rule" and Calculation 6.17, " Determination of Level 2 PRA Safety t Significant SSCs for the PNPP Maintenance Rule Program,"were reperformed to provide risk importance measures as the basis for establishing risk significance for l those systems modeled in the Probabilistic Risk Assessment and Probabilistic Safety l Assesment. The resulting risk importance measures, risk achievement worth, risk

reduction worth, and top 90 percent of core damage frequency cutsets were analyzed i

using method B of section 9.3.1 of NUMARC 93-01. The second step of that method

{ involves the elimination of cutsets or events that were not specifically related to

} maintenance (e.g., operator error and external or Initiating events). On reviewing the calculation, the inspector identified that cutsets and events involving common cause had also been remove On evaluation, the licensee determined that elimination of common caus3 cutsets and events could not be justified. Therefore these cutsets must be put back into listing and the cutoff points redetermined. This could result in additional functions being identified and reclassified high safety significance. The impact on the final result had not been determined at the close of the inspection, although the licensee expected to complete the reevaluation by September 5,1997. Consequently, this item could not be completely reviewed nor close With regard to the specific concems identfied in the report, the following was identified:

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P44 Turbine Buildina Closed Cooling Water The licensee considered this a low cafety significance system, ths initial assessment was based on an operating condition (2 of 3 pumps) which was not realistic in evaluating core damage frequency. In that scenario the contribution of the system was as an initiator and was at the bottom of the top 90 percent importance measure. The system contribution was reevaluated for the configuration it would be in during an accident mitigation scenarlo. The calculated contribution to core damage frequency was no longer within the top 90 percent. The licensee concluded that Turbine Building Closed Cooling Water was loss safety significant than the main turbine which was classified as low safety significanc C11 Control Rod Drive, Function C11 10 Reactor Pressure VessellevelInstrument reference leg fill The licensee considered this a low significance function. The justification for this classification v.as that the plant has four redundant reference legs and could accomodate the failure of one or two reference legs. Operators were trained to identify level notching phenomenon, which was frequently associated with volding in the Instrument reference leg, and trained to respond in accordance the plant's Emergency Operating Procedures. The licensee's Conduct of Operations procedure placed high priority on repair of the backfili subsystem if it was out of servic B21 Nuclear Galler, function B21 10 The licensee initially considered this as a low safety significance function. The justification for retaining that classification considered that only one of the four main steam lines needod to be open to provide necessary flow, that only one of the main steam shutoff valves needed to be open, and that this method was redundant to the opening of bypass valves which was a high safety significance functio M56 - Hydrogen Ignitor The licensee initially considered this a low safety significance system. The justification for retalning this classification was an Individual Plant Examination sensitivity analysis

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which assumed a complete failure by operators to put the hydrogen ignitors in servic Sensitivity changes with respect to No Reactor Pressure Vessel Failure, No Containmont Fallure, or No Pool Bypass Failure were 0 percent for all three, which indicated that operation of the Hydrogen Ignitor system was not significant. Setting reliability of the Hydrogen Ignitor system to 1 (perfect reliability) resulted in no change to No Reactoi Pressure Vessel Failure and No Containment Failure, and only a percent improvement for No Pool Bypass. Based on this minor change for a perfectly reliable system, the licensee concluded that the Hydrogen Ignitor system should remain as a low sa'ety significance system

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T23 - Containment Design, Functions 23-01,23-04,23-05,23-06,23-08 The licensee had originally provided no basis for the low safety significance i determination. The justification for retalning this classification was based on a sensitivity analysis of the Level 2 Probabilistic Risk Assessment with respect to No Reactor Pmssure Vessel Failure, No Containment Failure, or No Pool Bypass Failure. The analysis demonstrated that functions T23-01, T23 04, T23 06, and T23-08 were redundant to functions T23-02, T23-03. On this basis the Expert Panel concluded that .

classifying these functions as low safety significance was appropriate. The Expert Panel I concluded that function T23 05 for operation of the drywell automatic isolation valves was a high safety significance function although the Level 2 Probabilistic Risk Assessment did not identify the function that wa The inspector concluded that these justifications were reasonable.

M8.2 (Cloted) Insoection Followuo item 96014-03(DRS): Reliability / Unavailability Balance The inspector reviewed portions of the Cycle 5 periodic assessment report that addressed optimizing availability and reliability for structures, systems, and compor$ent The evaluation examined the performance of those structures, systems, or components which exceeded milability or unavailability criteria. Only one system (Reactor Core Isolation CoolinD, was noted as having exceeded unavailability criteria, primarily as a result of a maintenance effort to improve reliability. Subsequent performance revealed that tHs effort was successful. The analysis included a sensitivity study demonstrating that although unavailability of the Reactor Core Isolation Cooling system was greater than anticipated, unavailability of the remaining piant systems that was better than the criteria resulted in an 11 percent overall reduction of core damage frequency. The methodology used in the Cycle 5 assessment was considered acceptable and the conclusions were reasonabl Efforts to quantify the reliability / unavailability balance were still in progres Expectations remain that this quantative methodology will be used for the Cycle 6 assessment. Cycle 6 will be completed at the start of refueling outage RF06, which will begin on September 12,1997. The Cycle 6 assessment will be completed subsequent to RF06. Because the methodology and results from Cycle 5 were sufficient to meet the Maintenance Rule, and the quantitative method was an enhancement, the inspector concluded that further inspection was unnecessary. This item is closed.

M8.3 (Closed) Violation 96014-04(DRSP Failure to Perform Risk Assessment This item involved the licensee's failure to evaluate the risk significance of an emergent work activity during a Division 1 outage as required by a Perry procedure. To prevent recurrence of this event, the Perry on-line maintenance risk assessment process was significantly revised. Perry Administrative Procedure 0130,"On line Schedule Development," was revised to clarify respons;bilties, add additional review responsibilities, specify Maintenance Rule activities, require senior management approval for high risk configurations, and delete the 48-item risk list. An additional, new

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procedua, Perry Administrative Procedure 1924, Revision 0, dated August 7,1997,

• On-line Risk Assessment," was issued. This procedure identified respons:bilities and provided specific Instructions for performing risk assessments for emergent work situations and work not completed from a previous schedule. Instructions were also provided for incorporating risk assessment in the routine work scheduling process. A second new procedure, Plant Data Book C0011, Revision 0, dated July 23,1997, *0n-line Core Damage Frequency," was issued which identified systems that had little or no contribution to core damage frequency, listed pre analyzed conditions, and defined the four risk categories with respect to spans of core damage frequencies: risk category 1 for greater than 5.0X10d; risk category 2 for core damage frequency greater than 1.0X104 and less than or equal to 5.0X10d; risk category 3 for core damage frequency greater than 2.0X104 and less than or equal to 1.0X10d; risk category 4 for core damage frequency less than 2.0X104 This item is close M8.4 (Closed) Unresolved Item 96014 05(DRS): Reliability Performance Critoria Calculation 0.20, Revision 0, * Recommended Performance Criteria for Systems /1 rains Based on PSA Failure Rates," was reperformed to provide the basis for selecting performance criteria. In this sensitivity study, the number of expected demands was determined from test and surveillance schedules and the unit logs. The failure rate for each event, based on reliability criterion was quantified, and these failure rates were used in place of the base case failure rates. The core damage frequency was then recalculated with a result of 8.9X104 As a result of this study, the reliability criteda were redetermined in the previous selection the reliability criteria were established for the structures, systems, and components; in the current selection re!iablity criteria were also assigned to functions, in general, reliability criteria for structures, systems, and components remained the same. Some of the newly established reliability criteria for functions were more restrictive than the overall system reliablity critieria. The typical case was that a function would be allowed one failure over two cycles while a system might be allowed two failures over one cycle. Given that a system could have as many as a dozen functions, this was acceptabl Because the core damage frequency result from the sensitivity study was reasonable and there were only a few changes to the reliability criteria hr the structures, systems, and components, the licensee's reliability performance crite.ia were considered acceptable. This item is close M8.5 (Closed) Insoectica Followuo Item 96014-06(DRS): Structure Monitoring Program This item involved the need to establish appropriate and defined criteria for reclassifying structures from category (a)(2) to (a)(1) for monitoring. A review of TAl 0513, Revision ,

• Monitoring the Effectiveness of Maintenance, Structure Mon'toring Program," revealed that the definition of " unacceptable" now clearly required catargorizing the affected structure as (a)(1) until such time as the deficiency was corrected. Composition and qualifications for walkdown teams were specified. Guidance was provided to waNdown

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teams to assist in identifying deficiencies in structural steel, concreto, masonry walls, inaccessible, and miscellaneous items. This included, but was not limited to, aspects such as corrosion, concrete cracking, spalling and scaling, unfilled drilled holes, water ,

leakage, settlement, seismic gaps, and penetration seals. Based on the improved l guidance incorporated in the r6<lsed procedure, the licensee's structuro monitoring i program was acceptable. This item is close M8.6 (Closed) Insoection Followuo item 96014-07(DRS): Corrective Actions from Problem Identification Form 96 3276 Problem Identification Form 96-3276 was written to document the failure of the hydramotor on vortex damper OM25 F260A. The documentation associated with Problem Identification Form 96 3276 was reviewed to determine what corrective actions were being taken from the Malntenance Rule pere,pective. The Problem Identification Form concluded that the failure of the damper hydramotor resulted in a functional failure of the the Control Room Heating, Ventilation, and Air Conditioning system. This functional failure did not result in exceeding the system's reliability criteria, thetofore the system remained classified as category (a)(2). Corrective actions included an immediate rebuild of the hydramotor to replace a pinched o-ring and a 'eview of hydramotor maintenance and testing practices. A further evaluation of system operation was conducted when damper OM25 F260A failed again on July 28,199 Problem identification Form 97-1221 identified this as a repeat failure; however, the operability determination identified that failure of the damper would not provant the Control Room Heating, Ventilation, and Air Conditioning or the Control Room Emergency Recirculation systems functions from being fulfilled. The hydramotor failures documented under these two Problem Identification Forms were subsequently reclassified as component failures. Corrective actions from these two Problem Identification Forms remain ongoing. The inspector conciuded that although these failures did not change the systems' Maintenance Rule status, appropriate corrective actions were proceeding through the licensee's corrective action reporting progra This item is close L Management Meetinga X1 Exit Meeting Summary The inspector discussed the progress of the inspection with licensee representatives on a daily basis and presented the inspection results to members of licensee management following the conclusion of the inspection on September 28,1997. The licensee acknowledged the findings presente The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. None were identifie _ _ _ _ _ . _ . _ _ - _. .- . _ _ _ _ _ . ___ _ _ . _ . _ _ _ _ _

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PARTIAL LIST OF PERSONS CONTACTED LICRMita P. Rordley, Nuclear Engineering Supervisor K. Cimorelli, Scheduling Supervisor W. Colvin, Maintenance Rule Project Engineer D. Gudger, Licensing Engineer S. Meyer, PRA Engineer UBC D. Kosloff, Senior Resident inspector, Rlli J. Clark, Resident inspector, Rlli LIST OF INSPECTION PROCEDURES USED

'IP 62002 Inspection of Structures, Passive Components, and Civil Engineering Features at Nuclear Power Plants IP 62706 Maintenance Rule IP 90902 Maintenance Followup ITEMS OPENED, CLOSED, AND DISCUSSED 50-440/96014-02(DRS) UNR Potential misclassification of risk significant functions 50-440/96014-03(DRS) IFl Reliability / Availability balance 50-440/96014-04(DRS) VIO Failure to perform risk assessment 50 440/96014-05(DRS) UNR Reliability performance criteria 50-440/90014-06(DRS) IFl Structure monitoring program 50 440/96014-07(DRS) IFl Corrective actions from Problem Identification Form 96-3276 LIST OF DOCUMENTS REVIEWED Procedures

. TAl-0513, Revision 1, dated August 19,1997, ' Monitoring the Effectiveness of

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Maintenance Structure Monitoring Program *

• PAP-0130, Revision 1, Change 3, dated August 7,1997,"On-line Schedule Development'

. PAP-1924, Revision 0, dated August 7,1997,"On line Risk Assessment'

• PDB-C0011, Revision 0, dated July 23,1997,"On-line Core Damage Frequency"

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i Calculations

j e Calculation 6.16, Revision 1, ' Determination of Level 1 PSA Safety Significant SSCs for

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Calculation 6.17, Revision 1, * Determination of Level 2 PRA Safety Significant SSCs for the PNPP Maintenance Rule Program"

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Calculation 6.20, Revision 0, * Recommended Performance Criteria for Systems / Trains Based on PSA Failure Rates" i '

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Expert Panel Meeting Minutes, Meeting No. 87, Revision 2, dated October 4,1996

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Expert Panel Meeting Minutes, Meeting No.110, Revision 3 dated August 26,1997

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Table Maintenance Rule Function, Performance Criteria, and Classification, Revision 1.11, dated October 31,1996 1 .

Table M?.intenance Rule Function, Performance Criteria, and Classification, Revit son l 2.031, dated August 26,1997 4 .

Maintenance Rule System Report, Cycle 5, System: 99, Miscellaneous Structures,

approved 8/23/96

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