IR 05000440/1998014

Insp Rept 50-440/98-14 on 980620-24 & 0720-22.No Violations Noted.Major Areas Inspected:Licensee C/A for Issues Identified by NRC Inspections & Licensee Event Repts Submitted to NRC

ML20237C383
Person / Time
Site:	Perry
Issue date:	08/19/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20237C377	List: IR 05000440/1998014 ML20237C375
References
50-440-98-14, NUDOCS 9808210123
Download: ML20237C383 (12)

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U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket No: 50-440 License No: NPF-58 Report No: 50-440/98014(DRS)

Facility: Perry Nuclear Power Plant Location: 10 Center Road Perry, OH 44081 Dates: June 20-24 and July 20-22,1998 inspector: J. H. Neisler, Reactor inspector Approved by: R. N. Gardner, Chief Engineering Specialist Branch 2

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i EXECUTIVE SUMMARY l

Perry Nuclear Power Plant ,

NRC inspection Report No. 50-440/98014 l l

This inspection reviewed the licensee's corrective actions for issues identified by NRC inspections and licensee event reports submitted to the NRC. The inspector determined that the licensee was effective in identifying and implementing corrective actions to notices of violation, unresolved items, inspector follow up items and licensee event report ,

Root cause evaluations were thorough, accurate and well documented. (All Sections) l Staff involvement in the corrective action process was good. (All Sections)

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Ill. Engineering This inspection involved the review of the licensee's corrective action relative to NRC identified issues and licensee identified issues as reported in licensee event reports (LER). The inspector reviewed licensee responses, safety evaluations and interviewed cognizant licensee personnel.

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E8 - Miscellaneous Engineering issues

E (Closed) Licensee Event Report 50-440/93021-01. The improper Setting of Motor Operated Valve Results in Loss of Emergency Closed Cooling System Safety Function r

and Condition Prohibited by Technical Specification (Tech Spec).

On December 23,1993, the licensee identified excessive leakage through valve  !

OP42-F295A caused by a combination of personnel error and inadequate procedural l direction for setting motor operator valve (MOV) limit switches and mechanical stops.

I Limit switch and mechanical stop adjustment had been pcrformed on valve

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OP42-F295A on March 19,1993, according to the appropriate maintenance procedure in effect at that time. Valve retest consisted of only valve stroking and position verification. No post-maintenance leak test was performed. On July 1,1993, leakage

! through valve OP42-F295A was determined to be approximately 250 gallons per minute (gpm). Allowable emergency closed cooling (ECC) Loop A leakage was 3.3 gpm. As a result, from March 19 to July 1,1993, ECC Loop A could not perform its safety function ;

without reliance on division 2 components and systems being operable.

i The licensee's corrective action was to readjust the limit switches and mechanical stop Allowable leakage criteria was established and a post-maintenance leak test .

successfully performed. The procedure for Limitorque limit / torque switches was revised I to provide irnproved direction for setting limit and torque switches and adjusting ,

mechanical stops. The procedure revision also added post-maintenance test I requirements for butterfly valves that have established leakage criteria. Training was j provided to maintenance and engineering personnel on this issue including verifying i proper butterfly closure. In addition, the licensee has recategorized the valve as an ASME,Section XI, Category A valve. As such, the valve will be periodically tested

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undi the inservice testing program (IST). The ECC leakage and its effect on system i operability was identified as a violation in inspection Report 50-440/96008 during the i review of the previous revision of LER 93-021. This LER is considered close '

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E8.2 (Closed) Violation 50-440/97008-07. Failure to update final safety analysis report. The l licensee failed to update the Final Safety Analysis Report (FSAR USAR) to reflect plant i

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The inspector verified that the licensee had initiated change requests (CR)97-102 and 97-103 to implement the changes to USAR Table 3.9-30 and CR 97-068 to change Table 9.2-18,-19 and Figure 9.2-3. Change requests were in process to implement revisions to Table 8.3-1 by October 10,1998. Based on the above, this violation is considered close E8.3 (Closed) Licensee Event Report 50-440/92-005-01. Local Leak Rate Tests Result in Exceeding Technical Specification Failure Criteri During the third refueling outage (RF03), four containment penetrations exceeded the Local Leak Rate Test (LLRT) failure criteria. The causes for three of these events were component failures. The cause of the fourth event was attributed to misapplication of the check valve for that functio Penetration P306: The cause of the instrument air supply header containment inboard check valve failure was identified as leakage between the seat insert and the valve body. The seat insert is threaded into the valve body and torqued in accordance with  ;

the manufacturer's instructions. However, leakage developed between the threaded '

areas. The leak was repaired by replacing the valve intemals and sealing the threaded areas with grafoil tap Penetration P423: Investigation revealed that the main steam line drain and bypass isolation valve was not seated when closed electrically and the torque switch could not be adjusted to allow the valve to seat properly. The torque switch was replaced and the valve successfully teste Penetration P121: Failure of feedwater check valves 1821-F0032A and 1N27-F0559A was attributed to wear of the seating surfaces of the valves as part of normal operatio The valves were cleaned and seating surfaces lapped. The licensee initiated a trending '

program for use in determining the appropriate frequency for check valve maintenanc Penetration P204: Failure of the inboard containment isolation valve was attributed to the incorrect use of a check valve type which is susceptible to leakage at low pressur !

The licensee replaced the valve with a spring loaded lift check valve during RFO ]

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The inspector determined the licensee's corrective actions in regard to this LER to be appropriate. This LER is considered close l E (Closed) Licensee Event Report 50-440/97-015-00. Containment Penetration Leak Rate Test Failur A nuclear closed cooling (NCC) system containment penetration "as found" leakage i i exceeded LLRT failure criteria. The penetration could not be pressurized. The failure occurred because valves 1P43-F0721 and 1P43-F0055 were unable to maintain test pressure. The licensee's evaluation determined that 1P43-F0721 failed to maintain >

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LLRT pressure due to corrosion and pitting on the valve body seat to an extent that the

- O-ring on the valve disk was unable to form a'1 adequate seal. Corrosion and pitting occurred due to elevated oxygen levels in the NCC syste i

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Valve 1P43-F0055 failed to maintain LLRT test pressure due to an inadequate seal between the seat and disk. The valve seat, made of ultra high molecular weight polyethylene, had worn enough to prevent adequate sealing. The inspector verified that the licensee had repaired both valves and that both valves passed the post maintenance LLRT. Chemistry procedures were revised for hydrazine controls; monthly analyses were added for oxygen and suspended iron oxide; and quarterly analyses added for iron, copper, chloride and sulfate. The inspector concluded the licensee's corrective actions to be adequate. This LER is considered close E8.5 (Closed) . Violation 50-440/96004-03. Failure to initiate adequate corrective action to correct problems with the plant underdrain system. The licensee had previously _

identified calcium deposit buildup on pumps, valves, and instrumentation in the plant underdrain system. Scheduled tests and surveillance had not been performed as require The licensee has established a set schedule for the underdrain system. Tests to date have been completed as scheduled. The 9795 form," Justification of Non-Perforniance,"

Attachment 4, was revised to require a higher level of management approvals for successive / excessive test and surveillance postponements. This violation is considered close E8.6 ' (Closed) Inspector Follow up Item 50-440/93011-05. Type B testing of the inclined Fuel

- Transfer System (IFTS) Bellows may not identify 100 percent of the leakage from the two-ply bellows. The IFTS bellows (F42) was identified by the licensee as the only two-ply bellows matching the characteristics described in NRC Information Notice 92-2 The licensee initiated Design Change Package (DCP)93-157 to modify the IFTS bellows by installing an extemal boot between the bellows and the guard pipe and pressurizing the cavity. Type B test renuits subsequent to DCP-157 installation have been satisfactory. This issue is cons'dered close E8.7 (Closed) Licensee Event Report 50-440/94-004-01. Containment Penetration Local Leak Rate Testing (LLRT) Results Exceeding Technical Specification (Tech Spec)

Limits.-

The licensee reported that during the fourth refueling outage, two containment penetrations exceeded the LLRT failure criteria. This condition was identified during routine surveillance testing and during an operational condition (refueling) to which Tech Spec limiting conditions for operations (LCO) did not appl The licensee's corrective action was to repair the valves and perfum a LLRT to assure that the valve leakage met Technical Specification reWrements. The maintenance procedures and foreign material exclusion program were enhanced to prevent recurrence. The valves were again successfully tested during subsequent refueling outage RFO l L

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E8.8 (Closed) Unresolved item 50-440/93020-02. Potential design deficiency affecting the interface between the emergency closed cooling water (ECCW) system and the nuclear closed cooling (NCC) syste The licensee performed analyses and calculations that determined that the NCC piping met seismic design criteria that would assure pipe integrity during a seismic event. This precludes the drain down of the ECCW control complex chiller piping. Appropriate MOV stroke times have been established for the ECCW/NCC interface. The licensee also perfomied a review of plant systems to identify and evaluate safety /non-safety related system interface In addition, a USAR change was submitted to clarify the ECCW/NCC MOV stroke time Based on the inspector's review of the above, this item is considered to be resolve E8.9 (Closed) Licensee Event Report 50-440/97005-00. Automatic Reactor Scram Following Auxiliary Transformer Tri The plant experienced an automatic reactor scram June 5,1997, as a result of a fault in the 13.8kV cable termination compartment. The fault was a phase A fault-to-ground on bus L12 caused by moisture and dust buildup on the bus insulator / support. As a result i of the Phase A ground, ionized gases, soot, dust and debris faulted Phases B and l

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The transient caused the auxiliary transformer 110-PY-B to isolate on a phase differential signa Bus lockout on an actual fault condition was expected but the lockout occurred on bus I L11 instead of on the faulted bus, L12. The licensee's investigation found that I protective relaying for non-safety buses L11 and L12 had been reversed. The reversal I was determined to be the result of improper construction wiring, i

The licensee's root cause analysis determined the cause to be a missing piece of gasket j material on the termination box housing that permitted precipitation to collect in the compartment. After abnormally rainy weather, followed by a period of dry weather, the combined effect of moisture and dust created a path to ground across the insulator to the support. The missing gasket section was a small spliced section approximately two inches by two inches on the termination compartmen The licensee repaired the 13.8kV bus bar terminations, replaced tennination compartment gaskets, performed auxiliary transformer testing which determined that no intemal damage had occurred, and corrected the wiring for protective relaying on buses L11 and L12. In addition, training to assure electrical equipment is adequately protected was conducted for appropriate maintenance personnel. This LER is considered close E8.11 (Closed) Severity Level IV violation 50-440/97008-09. The licensee had modified j various electrical systems as reflected in design drawings and had not revised nor j updated the design calculation _ _ _ _ _ _ _

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The licensee initiated Problem Identification Forms (PIFs) to address inconsistencies between data in calculations and data in design basis documents, USAR Chapter 8 and design drawings. The inspector reviewed the PlFs including the calculations associated with the specific issues identified in the violation. Licensee reviews performed under

' PlF 97-0517 identified 59 safety related electrical calculations that had not been reviewed within the past five years. Reviews were initiated and 42 of the calculation reviews have been perfonned and revised as necessary. The remaining 17 calculation reviews were in process. This issue is close l E8.12 (Closed) Licensee Event Report 50-440/94-015-01. Potential Loss of Emergency

- Service Water System (ESW) Due to Loss of Keepfil On June 9,1994, operators observed that ESW keepfill pressure was below the required pressure. Review of the keepfill system identified a deficiency in a recent r

service water (SW) design change which reduced keepfill pressure under certain operating conditions. Immediate corrective action was to transfer the keepfill system to its attemate source. An engineering review determined that during periods of Lake Erie low water temperature, ESW system operability could be affecte ;

Since an attemate source, P54 Fire Protection, was available to maintain keepfill 4 pressure, and adequate pressure was maintained to prevent in-leakage, safety significance of this event was minima The licensee developed and installed a design change to effect modifications to assure adequate water to maintain keepfill system pressure to the ESW sts.idpipes. The design change involved installing stainless steel piping from a service water tie-in point upstream of the turbine tube oil coolers in the turbine building to the top of ESW standpipes in the auxiliary building. The design change was completed in March 199 The LER is considered closed.

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E8.13 (Closed) Licensee Event Report 50-440/94-014-01. Failure to Meet Technical Specification (Tech Spec) Limits for Main Stream Line Leakage Rate !

During investigation of local leak rate testing (LLRT) results on the main steam line, the licensee determined that the testing methodology used was not representative of design

! basis accident conditions. - Therefore, the LLRT did not satisfy the surveillance requirements of the Tech Specs. The licensee attributed the root cause to inadequate

procedures. The procedures allowed nonconservative testing of the main steam isolation valves (MSIV) since additional seating force was allowed to be applied to the MSIVs by non-safety-related instrument air. The cause of excessive leakage was seat l alignment on three MSIVs and packing leaks on the othe On June 16,1994, the licensee met with NRC staff at NRC Headquarters to discuss the status of the MSIVs and licensee action to assure MSIV integrity. The NRC staff concluded that the licensee's plans and activities were acceptable for plant operabilit (

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I The inspector reviewed the.results of the licensee's investigation, corrective actions and i results of the retest using safety related instrument air. In addition, the inspector i

rwiewed CR 94-146, training documentation, drawings / sketches for safety related instrument air to MSIV modification and CR 940166. The inspector determined that the 1 corrective actions were acceptable. This LER is close ]

l . E8.14 (Closed) Enforcement Action (EA)97-430. The licensee changed the emergency l closed cooling (ECC) surge tank sizing from a seven-day supply to a 30-minute supply, l with operator actions required outside the contro! room to initiate makeup from the emergency service water (DSW) system. The NRC concluded that this change constituted a potential unreviewed safety question as defined in 10 CFR 50.59 since the probability of occurrence of malfunction of equipment important to safety was increased.

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The licensee's corrective actions appeared reasonable and adequate. The ECC i L- boundary valves were leak tested and the ECC system verified to be within USAR limits.

l The ECC system configuration was retumed to its design / licensing basis, in addition, the licensee performed a review of the safety evaluation process that included an independent review of selected outage safety evaluations. Training of personnel included Plant Operational Review Committee (PORC) members. This EA is

considered closed.

E8.15 (Closed) Licensee Event Report (LER 50-440/95-003-00) Loss of Safety System Function Caused by inoperable RHR System Snubbers.

! The licensee reported that all three trains of the Residual Heat Removal (RHR) system I

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may have been inoperable during periods between initial plant startup and the beginning of the fourth refueling outage (RFO4). Two of the trains were technically inoperable because of missing swell deflectors on snubbers located within the swell pool regio During periods when the third train of the RHR system, train A, was out of service or inoperable, all three trains would be considered inoperabl As a result of design changes to suppression pool swell loading elevations during  !

, 1980-1983, the suppression pool swell elevation was raised from 620 feet 6 inches to L 623 feet four inches. Walkdowns of the suppression pool snubbers in 1984 to l determine acceptability of snubber locations only considered small bore piping snubbers I

since at the time it was considered that the large bore piping had sufficient controls to assure snubbers on large bore piping would be above the suppression pool swell froth

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line. The licensee initiated Condition Report 94-955 to determine whether swell shields L were required. Their investigation determined that shields were required and design -

details were provided for installing swell shields on the three snubbers. No other

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snubbers requiring shields were identified uuring the licensee's investigatio ;

( The licensee's evaluation determined that the three snubbers had been inoperable in  !

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Modes 1 and 2 from hitial startup until the plant entered Mode 3 at the beginning of RF04. The assumption was that the snubbers mcy lock up after the first actuation when

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submerged by the suppression pool swell. Discharge to the pool would not occur in i Modes 3,4 and 5; therefore, the snubbers were considered inoperable only in Modes 1 and , The licensee installed pool swell shields on the snubbers prior to startup following RF0 T he snubber are now considered to be operable in all plant operating modes.

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The failure to identify and correct the missing snubber swell shields during the previous walkdown and design reviews that resulted in the RHR system being inoperable is

! considered a violation of 10 CFR 50, Appendix B, Criterion XVI," Corrective Action."

This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Polic (50-440/98014-01)(DRS)

V. Management Meetings X1 Exit Meeting Summary The inspector presented the inspection results to members of the licensee management at the conclusion of the inspection on July 22,1998. No proprietary information was identified by the license i

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PARTIAL LIST OF PERSONS CONTACTED Licensee

' H. Bergendahl, Director, Nuclear Services R. Collings, Manager, Quality Assurance T. Henderson, Complianca Supervisor J. Messina, Manager, Maintenance Support L. Myers, Vice President- Nuclear J. Powers, Manager, DES S. Sanford, Compliance R. Schraudar, Director, Engineering M. Wesley, superintendent, Maintenance NBC D. Calhoun, Senior Resident inspector J. Clark, Resident inspector INSPECTION PROCEDURES USED IP 92700 Licensee Event Reports IP 92702 Follow up on Corrective Actions for Violations and Deviations

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I ITEMS OPENED, CLOSED AND DISCUSSED l l

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50-440/98014-01 NOV Corrective Action Closed 50-440/98014-01. NCV Corrective Action  !

50-440/94014-01 LER MSIV Leak Rate Testing 50-440/97-43 EA ESW Surge Tank Sizing 50-440/94015-01. LER ESW Keepfill Pressure i 50-440/97008-07 VIO FSAR Update  !

50-440/92005-01 LER LLRT Multiple Containment Penetration Failure 50-440/97015-01 LER LLRT NCC Penetration Leak 50-440/96004-03 VIO Underdrain System Testing I 50-440/93011-05 IFi inclined Fuel Transfer Bellows 50-440/95003-00 LER RHR Inoperable Due to Snubber Swell Deflectors Missing 50-440/93020-02 URI Open/Close Timing of NCC System Values

'50-440/94004-01 LER LLRT Penetration Failure 50-440/97008-09 VIO Control of Electrical Calculations i 50-440/93021-01 LER ECC Value to NCC isolation Leak 50-440/97005-00 LER ' Scram Following Auxiliary Transformer Trip l-50-440/97007-04 IFl Scram Following Auxiliary Transformer Trip . !

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LIST OF ACRONYMS USED CFR Code of Federal Regulations CR Change Report DBE Design Basis Earthquake DCP Design Change Package ECC Emergency Closed Cooling ESW Emergency Service Water GPM Gallons Per Minute IFTS Inclined Fuel Transfer System LCO Limiting Condition for Operation LER Licensee Event Report LLRT Local Leak Rate Test Loop Loss of Offsite Power MOV Motor Operated Valve MSIV Main Steam isolation Valve NCC Nuclear Closed Cooling PlF Potential issue Form PORC Plant Operational Review Committee RFO Re-Fueling Outage RHR Residual Heat Removal Tech Spec Technical Specification

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USAR Updated Safety Analysis Report i

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