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Insp Rept 50-293/97-13 on 971111-980106.Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support
	ML20202D254
Person / Time
Site:	Pilgrim
Issue date:	02/06/1998
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20202D240	List: IR 05000293/1997013; ML20202D237; ML20202D244;
References
	50-293-97-13, NUDOCS 9802170027
	Download: ML20202D254 (81)
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Enclosure 2 U.S. NUCLEAR REGULATORY COMMISSION-

REGION I

License No.: DPR-35 *

Report No.: 97-13 Docket No.: 50 293 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station

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inspection Period: November 11,1997 January 6,1998 Inspectors: R. Laura, Senior Resident inspector R. Arrighi, Resident inspector D. Dempsey, Reactor Engineer E. Connor, Project Engineer Approved by: Curtis J. Cowgili, Ill, Chief Reactor Projects Branch No. 5 Division of Reactor Projects

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EXECUTIVE SUMMARY .

Pilgrim Nuclear Power Station NRC Inspection Report 50-293/97 13 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers resident inspection for the period of November 11,1997, through January 6,1998; in addition, it includes the results of announced inspections by a regional MOV specialist and a regional projects inspecto Ooerations I

=+ 'Iwo reactor startups were characterized by clear communications and effective control by shift supervision. Senior reactor operator performance was good in that '

distractions to the operating crew were minimized, and the diverging vessel to flange temperature differential was identified and the plant heatup secure (Section 01.1)

Three operator human performance issues were evident. BECo identified that a reactor operator did not detect a diverging temperature trend between the reactor vessel flange and the adjacent shell temperature. BECo also identified that an inadvertent RWCU isolation during post scram conditions was an operator performance issue that broadly warranted increased training on the use of the letdown valve CV 1239. The NRC identified a partially clogged cooling air inlet screen on the "A" RHR pump motor which was not detected by operators on rounds or managers on tour. (Sections 01.1 and 02.1)

Operators responded effectively by using proper command-and control and procedure usage in response to a feed water system regulating valve malfunction which resulted in a turbine trip and resultant automatic reactor scram. The post trip review completed by operations support personnel and the readiness for restart meeting focused on proper evaluation and resolution of any equipment and human performance issues. (Section O2.1) >

Maintenance

Routine and emergent maintenance tasks were completed by experienced workers with generally good results. A planned emergency diesel generator LCO maintenance outage was better planned and executed than one during the previous period. (Section M1.1)

+= The NRC identified two l&C human performance issues with the potential use of expired thread sealant and the inadvertent cp.ning of a petcock valve on the lube oil strainer for the "A" ED The ATWS system was declared inoperable when l&C technicians inadvertently installed a 24 vdc relay into a 125 vdc apalication which was also considered a human performance issue. The relay overheated when placed in service, disclosing the human error during the activity. The event was caused by a work planner and an l&C supervisor, who changed the status of the work package from "on hold for parts" to " task ready," when, in fact, all parts were not available. Technicians in ii

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the field identified discrepancies with the replacement relay, but were incorrectly informed by the supervisor and a system engineer that the relay was acceptabl These errors resulted in a violation of the work control process requirement In three human performance related instances during this period, I&C supervisors missed opportunities to correct adverse conditions, and on one occasion this led to inoperable important to safety equipment. (Section M1.1)

e An unplanned shutdown resulted from degraded MSIV actuator closing springs which was considered a maintenance functional failure. BECo expanded the corrective actions to restore margin on all MSIVs. The use of a quarterly surveillance test, beyond the TS requirements, to detect actuator degradation at an early stage was a pos'tive initiative. (Section M2.1)

Maintenance troubleshooting efforts were effective in identifying the most probable cause why FV 642A failed open on December 6,1997, resulting in a reactor scra The BECo post trip review team determined that an l&C cc,rrective maintenance activity during the MSIV shutdown outage inadvertently interfered with the proper positioning of the valve clip located in the valve positioner. This was another l&C human performance issue. The system engineer properly applied the maintenance rule criteria for this maintenance preventable failure. Corrective actions were thorough. (Section M4.1)

Enaineerina

An unplanned power reduction and TS logging violation resulted from incorrect readings from a temporary modification made on the reactor vessel flange tempereture elements. BECo did not adequately evaluate the cause of and develop effective correc'Ive actions to preclude the recurrent instances of the temporary temperature detectors from becoming dislodged from the reactor vessel flange. The ( elements moved /3eparated from the reactor vessel flange on three separate

! occasions. The f ailure to adequately evaluate and correct the problem is considered i a violation. (Section E1.1)

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The licensee demonstrated the design-basis capability of its GL 8910 MOV However, the licensee's valve factor assumptions for some non differential-pressure tested MOVs need to be strengthened. BECo committed to obtain additional information to bolster the vaumptions by June 30,1998. The licensee appropriately adiJsted MOV program design assumptions in response to industry information. Program assumptions regarding load sensitive behavior, stem friction coefficient, and extrapolation of test data were acceptable for GL 89-10 program closure. (Section E1.2)

Based on available margins, the licensee demonstrated design-basis capability for each of the GL 89-10 MOVs at Pilgrim. (Section E1.3)

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+ in most cases, design + asis MOV calculations were updated to reflect the latest'

F technical information, isolated cases were identified in which calculations needed to be updated, but the licensee was aware of the discrepancies and revisions either were planned or in progress, (Section E1.4)

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BECo's tracking and trending program was consistent with the guidelines established in Generic Letter 8910. The licensee made effcctive use of industry information. (Section E2.1)

+ The licensees' MOV periodic verification program was not yet finalized as of the

- close of this inspection However, the GL 89-10 program is being closed because further NRC review of this matter will be conducted under GL 96 05. (Section E2.2)-

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With a single SSW pump operation per loop and under degraded power conditions,

, the SSW breaker overload heater protection relays for any pump could have tripped

the pump and, therefore, violate TS 3/4.5.8, Oper?le Core and Containment

-Cooling Systems. This condition was unknown by the licensee prior to receipt and >

review of the new SSW pump performance testing curve in August 1997. The licensee has determined that this condition was reportable. This condition is considered an apparent violation of design control per 10CFR50, Appendix B, Criterion Ill, (Section E2,3)

. Corrective actions have been taken by BECo to address the backlog of open vendor manual changes. A downward trend of open VMCRs has been noted with plans to

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address the older VMCRs. The open VMCR have not contributed to problems with plant equipment. The failure to process all vendor manual' changes in a timely manner was previously determined to be a non cited violation, No additional

violations of NRC requirements was noted; therefore, this item is closed, (E8.2)

Plant Sunoort

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Overall, proper radiological controls were demonstrated by the maintenance craft during replacement of the main steam safety relief valve. One poor practice was identified when a maintenance supervisor demonstrated poor contamination control work practices which had the potential to cause a personnel contamination by retrieving his cellular phone from inside of the anti contamination clothing without removing his rubber gloves and placing the receiver against his face,- (Section R4.1)

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TABLE OF CONTENTS E X EC U TIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . ll l Summary of Pla nt Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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l . O P E R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

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01.1 General Comments (71707) ...........................2

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O2 Operational Status of Facilities and Equipment ................... 3 02.1 (Closed) LER 50 293/97 26; Automatic Reactor Scram ........ 3 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 08.1 (Closed! Violation 50 293/97-01-01 and LER 50-293/97-03-01:

Procedure adherence to feed water system off normal procedure and '

loc ked valve list . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 11. M A I N T E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1.1 General Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 9 M2.1 (Closed) LER 50-293/9'7-25, Shutdown due to Inoperable Main Steam t

' isolation Valve (MSl\. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M4 Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . 10 M4.1 Feedwater Regulating Valve Maintenance troubleshooting . . . . . 10

M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . t 1 M8.1 (Closed) Violation 50 293/97-01-02: Inadequate corrective actions for three operator work arounds and also_for a missed standby switchyard battery surveillance . . . . . . . . . . . . . . . . . ...... 11 M8.2 (Closed) URI 50-293/96-80-02: Work Control issues . . . . . . .. 13 111. EN G I N E ERI N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E (Open) VIO 50-293/9713-02: Reactor Vessel Flange Metal Temperature indication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E1.2 (Open) (IFl 50 293/97-13-03/04/05); Generic Letter 8910 Motor-Operated Valve Program Review . . . . . . . . . . . . . . . . . . . . . . . 16 E1.3 Motor Operated Volve Design-Basis Capability ...... ..... . 21 E1.4 Design Calculations . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . 22 E2 Engineering Support of Facilities and Equipment ................. 24 E2.1 Tracking and Trending Program . . . . . . . . . . . . . . . . . . . . . . . . 24

. E2.2. Periodic Verification of MOV Performance .............,.. 24 E8 Miscellaneous Engineering Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 E (Closed) Violation 50-293/97-02-02: Inadequate Design Control . 27 E8.2 (Closed) URI 95-22-01: BECo Corrective Actions to Address Vendor M anual Upd ates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 v

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E8.3 - (Closed) URI 96-0102 Reactor Core Isolation Cooling (RCIC) System

tr >oerability (Closed) LER 50 293/96-02 . . . . . . . . . . . . . . . . . . 29

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. E8.4 (Cicso:0 LER 50 2 93/9 5 08 . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 E8.5 (Closed) LER 50 2 9 3/9 5 12 . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 E8.6 (Closed) UnresAd item 50 293/93 22 01: Pressure locking and

thermal binding of gate valves . . . . . . . . . . . . . . . . . . . . . . . . . 30

, E8.7 (Closed) Unresolved item 50 293/93-22-02: Ambient temperature effects on AC motors. .............................30 E8.8 Review of Updated Final Safety Analysis Report - . . . . . . . . . . . . 31 E8.9 Criticality Accident Requirements .....................31 IV. PLA NT S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 R1 Radiological Protection end Chemistry (RP&C) Controls ............ 32

R1.1' (Closed) Violation 97 03 02: Radiological Controlled Area (RCA)

Bou nd a ry D oo r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 R4 Staff Knowledge and Performance in RP&C ....................32 R4.1 Radiological Work Practice . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 V. M AN AG EM ENT M EETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 X3 Management Meet;og Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33

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X4 Review of UFSAR Commitments . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 33 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34

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li EMS OPENED, CLOSED, AND UPDATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 LIST O F AC RO NYM S U S ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7

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REPORT DETAILS !

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Summarv of Plant Status

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Pilgrim. Nuclear Power Station (PNPS) began the period at approximately 100% reactor

power. Significant power reductions of interest to the NRC are described below. --

'On November 14,1997, the failure of a local level control valve on the "D" moisture

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separator drain tank caused a high level condition on the tank. Operators lowered power level to 73% to effect repairs.- 100% power ~was restored on November 15,199 '

- Operators conducted'a normal plant shutdown on November 23,1997, when two main ;

steam isolation valves (MSIVs) did not pass the slow close surveillance test acceptance

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criteria. Further details of this event are discussed in section M2.1 of this report. A!ter

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performing corrective maintenance on the MSIVs,' operators brought the reactor critical on

. December 2,1997, and commenced power ascension.- r

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On December 5,1997, at approximately 21 % reactor power, power ascension was stopped and power reduced to support a drywell entry to resolve a 4 cubic foot / minut '

nitrogen leak. Operators found an MSIV accumulator drain valve with seat leakage until i the valva was torqued Jhut. - Power ascension was recommenced later that day.-

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, . December 6,1997, an automatic reactor scram occurred from 76%' reactor power. A feedwater system regulating valve malfunction resulted in a high reactor vessel water level

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causing a turbine trip and resultant reactor _ scram. A formal NRC notification was made per

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. Event #33360. Further details of this event tre discussed in section 02.1 of this report. ; ,

j: After repairs were made to the regulating valve, operators brought the reactor critical on ~ l

. December 9,1997. -100% reactor power was achieved on December 11 1997.

-: The "B" reactor recirculation system pump tripped unexpectedly on' December 12,1997 causing reactor power to decrease to approximately 75 power.. Operators subsequently lowered reactor power to approximately 35% to restart the "B" recirculation pump

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following completion of maintenance troubleshooting.100% reactor power was achieved ,

i on December 13, '1997. c l The plant operated at approximately 100% reactor power dunne the remainder of the -

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1. OPERATIONS 01 Conduct of Operations'

01.1 General Comments (71707)

Using inspection Procedure 71707,the inspector conducted frequent reviews of ongoing plant operations, in general the conduct of operations was professional and safety conscious. During tours of the control room, the inspector discussed alarms and abnormal conditions with operators to determine the significance and planned corrective action. FM example, the inspector observed operators promptly identify and respond to a data failure message (i.e., "D-Link Failure") on the digital controller for the "A" reactor recirculation pump. Operators followed the guidance contained in the alarm response procedure to manually lock-up the motor-generator set scoop tube. Additionally, a priority one maintenance request and two problem reports were generated in additior to promptly notifying engineering personne The inspector attended the plant manager's morning meeting where an emphasis was placed on evaluation and resolution of degraded equipmant conditions. The inspector observed the assistant operations department manager and plant manager tour the control room each mornin Operators closely monitored increasing temperature in the tailpipe from safety relief valve (SRV) 3B. The requirements of technical specification (TS) 3.6.D.3 and 3.6.D.4 for SRV seat leakage were well understood and followed. An operability evaluation was completed in advance of reaching the 212 degrees Fahrenheit limit. As a precautionary measure, the nuclear operating supervisor (NOS) briefed the crew on the immediate actions for a stuck open relief valv Operators exhibited effective control and procedure adherence during two plant start-ups and also in response to the unexpected trip of the "B" reactor recirculation pump. During the start-up activities, a designated reactivity manager was assigned tc provide augmented oversight of all control rod manipulations. The Nuclear Watch Engineer (NWE) closely monitored control room activities and minimized distractions to the crew by limiting unnecessary discussions in the control room and assigned a third operator to perform routine control room activities. The approach to criticality and subsequent heat up rate were conservative with respect to TS requirements. Additionally, reactor engineers interfaced smoothly with the operating crew on managing core reactivity. Also, operators promptly detected that control rod 2-23 was stuck and electrically disarmed the directional control valves pursuant to TS 3.3.A.b, Reactivity Limitation ' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _

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During the reactor start-up, BECo identified an operator performance issue involving the data acquisition and evaluation for the 145 degree differential temperature limit between the reactor vessel flange and adjacent vessel shell per TS 3.6.A.1, Primary System Boundary Thermal Limitation. A nuclear operating supervisor (NOS) identified a divercing temperature trend which was not detected by the reactor operator recording the data. This was identified during_ supervisory review prior to exceeding the 145 degree limit. As a result, operations department management briefed all crews on the importance of promptly evaluating operating data. Further details of the related degraded temperature elements are

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discussed in Section E1.1 of this repor Anomalies identified during tours by the inspectors were discussed with the NWE. For example, the inspector identified that the upper motor air inlet screens on the "A" residual heat removal (RHR) were dirty. The NWE initiated problem report (PR) 97.9008 to document and evaluate the condition. Later that day, the "A" RHR pump was run for one hour and all temperatures were in the expected band. The next day the air intake screen was removed from the motor, cleaned and r : installed. The inspector observed the air screens on the other ECCS pumps and found no similar problem Also, the inspector identified that snubber H31 134, which supports small bore piping for the west side scram discharge volume drain line, was missing a cotter pin in one of the snubber clevis pins. The NWE initiated PR 98.9009 to document and evaluate the l condition. Engineering personnel determined that the snubber remained operable as long l as the clevis pin remained in place. The missing cotter pin was installed by the end of this i

inspection perio The inspector discussed these two degraded conditions with the plant manager who noted that operator on rounds and managers on tour for the supervisory walkdown program missed an opportunity to identify these conditions. However, the inspector noted that overall numerous degraded conditions were identified by the plant staff and entered into L the corrective action syste Operational Status of Facilities and Equipment 0 IClosed) LER 50 293/97-26: Automatic Reactor Scram Inspection Scope (71707.93702)

.The inspector reviewed integrated crew and equipment performance following an automatic reactor scram that occurred on December 6,1997, due to an eqaipment problem. The inspector was in the control room at the time of the scram and directly observed o;tsrctor immediate actions to stabilize plant conditions. Additionally, the post scram review, readiness for reetart meeting and licensee event report (LER 97-26) were reviewe Further details of the maintenance troubleshooting and corrective maintenance actions are contained in Section M4.1 of this repor .. . . .

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l Observations and Findings 1 A high water alarm was received in the control room and 38 seconds later an automatic scram occurred due to a fai;ed open feed water regulating valve. Operators had observed the increasing water level and had attempted to %ke manual control of the regulating valve I prior to the scram. Operators stabilized plant coc.tlons in accordance with procedure 1.2.6, " Reactor Scram," and emergency operating procedures (EOPs). The inspector observed effective command-and control by the nucle,r operating supervisor (NOS). l During the normal post-scram " shrink" in vessel water level, isolation of the group 2 (drywell),' group 3 primary (RHR valves) and group C (RWCU) primary containment isolation control system (PCIS) functioned as designe Operators entered EOP-2, " Failure To Scram," due to 6 control rods witt.out fuil-in indication. Fifteen minutes after tha scrsm the control rod drive charging valve HO 30125 was closed and the 6 control rods resettled and the full-in lights functioned properly. The temporary loss of full-in position indication has occurred in the past and is discussed further in section 08.1 of this report. The inspector also noted that off-watch licensed operators in the control room provided assistance in carrying sut actions but only under direction from the NOS. Safety related equipment performed as designe l The BECo post scram review noted that approximately 11 minutes after the scram, operators were in the process of lowering reactor vessel water level by rejecting water through the RWCU letdown valve (i.e., CV-1239 when the RWCU system isolated. This group 6 isolation occurred due to a nonsafety-related function on high resin temperature at 140 degrees Fahrenheli. The reactor operator throttled open CV 1239 too far which resulted in increased letdown and increased resin temperature. The post trip reslew and operations depMment rmnagement identified this as a lessons learned and considered this an operator performance issue. Operations department management requested the

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operator training staff to enhance the training provided to the operators in this regard, in summary, operators stabilized plant conditions in a safe and expeditious manner following an automatic reactor scram caused by an equipment malfunction. The operating crew exhibited positive command-and comol when executing procedural steps of abnormal and EOPs. The resident inspector conducted an on-site review of licensee event report (LER) 97-26, dated January 5,1997, and determined that the report met the requirements of 10 CFR 50.73. LER 97-26 is considered closed. The BECo post trip review rigorously evaluated integrated equipment and operator human performanc After complethn of ma;ntenance troubleshooting and repairs to the ' ' feed regulating valve, the inspector observed the readiness-for-restart panel meeting lee by the plant )

manager. The inspector note: that there was substantial senior management support at the l meeting. Each discipline reviewed work completed and any deferred work that could affect reactor start-up. The restart panel also received a special briefing on the results of an !

independently chartered group which broadly reviewed plant equipment status. This review included items that had the potential to initiate a plant transient or delay start-u For example, the independent team identified that the organization had a tendency to be less aggrt ' :ive in resolving non-technical specification equipment issues where redunde-'

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e existed. The inspector determined that the restart panel conducted a thorough review and-that the initiative to initiate an independent team to review potential piant wide degraded

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conditions reflected a proper nuclear safety focus, Conclusions Operators responded effectively by using proper command and-control and procedure usage 1, response to a feed water system regulating valve malfunction which resulted in a turbine trip and resultant automatic reactor scram. The post trip review con.p.eted by operations support personnel and the readiness for restart meeting focused on proper

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evaluation and resolution of any equipment and human performance issue Miscellaneous Operations issues (92700,92901)

0 (Closed) Violation 50 293/97-01-01 and LER 50 293/97-03-01: Procedure ghetence to feed water system off-normal orocedure and locked valve list The NRC identified two examples where operators did not follow plant procedore instructions, in the first example, operators did not promptly manually scram the reactor due to a feedwater system regulating valve malfunction which led to difficulty in maintaining reactor vessel water level within the prescribed band Operators cycled the electric feed pumps on/off a few times to maintain reactor vessel water level to allow time for a condenser bay entry to attempt to take manual control of the feed water regulating valves. Operators were not trained to control reactor vessel water levelin this manner. In

, the response, dated April 17,1997, to the violation, BECo admitted to the violation and included the lessons learned in the operator training program. Also, BECo management expectations for adherence to off-normal procedures were reviewed with all plant op3rators. BECo review was also initiated of the clarity of the instructions contained in off-normal procedures to ensure procedure adequacy. Tha resident inspector conducted an on-site review of LER 50-293/97-03 01 and determined that the report met the requirements of t 0 CFR 50.7 The second example involved twa melted plastic valve restraints used in high temperatuse applications for two safety-related valves contrary to the procedural directions not to use non-metallic restraints. BECo admitted to the violation. Metallic restraints were immediately installed to properly restrain the two valves. Further, BECo conducted a review of other restraints for valves in high temperature applications to ensure metallic restraints were used. No other problems were identified during this review. This problem was included in operator training sessions. During this inspection period, the inspector randomly checked hi0h temperature valve applications and confirmed that metallic locking devices were use Based on the corre" 9 actions taken and the overall positive performance of operators during the NRC inspection of the 1997 licenced operator requalification examinations, the

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violation and LER 50-293/97-03-01 are close p a 48 -- -- J+. --< ,,i- L- u F -- J,- 4: --& --d-F

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11. MAINTENANCE M1 Conduct of Maintenance M1.1 General Maintenance insoection Scoce (61726,62707)

The inspector observed portions of selected maintenance and surveillance activities to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to limiting conditions for operation, and correct system restoration following maintenance and/or testing. Portions of the following activities were observed:

19301726 "A" EDG coolant level switch modification 3

19701028 "A" EDG frequency meter / transducer 6 19703034 Troubleshoet trip of "B" recirculation M/G set

19601062 Replaco pilct assembly on SRV 3D

19702989 Troubleshoot malfunction of the "A" feed water system regulating valve

8M1-3 Att 6 "B" and "D" APRM functional checks

P9501155 HPCI booster pump oil change preventive maintenance

19702761 Test operation of RHR logic test relay 10A K118 Observations and Findinas Work progressed as scheduled du-ing a planned LCO maintenance outage on the "A" emergency diesel generator (EDG). I&C workers installed root valves into the EDG coolant expansion tank and lifted the electrical leads for the old level switch in a local terminal bo Overall, workers were observed to be experienced and utilize good work practices. An l&C supervisor and work week manager were observed providing oversight of 16e various work tasks. The EDG outage went smoother than the one performed during the pravious'

inspection period. However, due to a parts restraint problem with the repla':ement level switches, only a portion of the level switch modification was completed. The work control manager informed the inspector that this element of the job involving parts would be specifically reviewed in the lessons learned report. The portion of work completed will allow completion of the modification at a later time without disabling the EDG for long '

periods of tim The inspector identibed two problems at the work site. The first problem involved a tube

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of thread sealant that was past the shelf life expiration date. The thread sealant had not yet been used during the work when identified. The l&C supervisor obtained a new tube of thread sealant for use during the work and initiated problem report (FR) 98.0003 to document, evaluate and develop corrective acti)ns for the problem, as needed. However, the fact that the wrong materials were issued for use in the field was a violation of the ,

licensee's work control process procedures. This failure constitutes a violation of minor significance and is being treated as a Non-Cited Violation, consistent with Section IV of the

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NRC Enforcement Pally (NCV 971312). While minot violations are not normally described in the inspection report, this issue relates to a more significant violation that occurred during this same inspection period involving relay maintenance that was cited, and is described later in this section of the inspection repor The second problem was identified by the inspector after the physical work was completed and preparations were in progress to perform a maintenance run of the EDG. Lube oil was observed leaking from the petcock vent valve B8-45006A on the lube oil straine Approximately one cup of lube oil leaked out. An operator preparing to start the EDG closed the petcock which was approximately 1/4 turn open. The inspector noted that l&C supervision present at the work site missed an opportunity to detect and correct these two probicms. The licensee wrote a problem report to evaluate the issu Also during 's inspection period, an l&C human performance related event occurred which rendered divleion 11 of the anticipated transient without scram (ATWS) system inoperable for a short pariod of time. On December 31,1997, I&C technicians inadvertently replaced a 125 vde Agastat relay with a 24 vdc relay as part of a preventive maintenance activit Shortly afterward, the relay overheated and caused an alarm in the main control roo Prior to installing the correct relay, BECo developed a test procedure to sequence the work and confirm proper operation of ATWS circuits. The inspector attended a multi-disciplinar/

meeting on December 31,1997, where a corrective action plan was developed. A proper safety focus was evident at the meeting based on the d!scussions of the interim status of the subject ATWS panel. Operators properiy entered the plant into a 14 day technical specification LC l No other damage was detected on the ATWS panel circuitry. The relay was replaced with

} one of the correct rating and the system was declared operable. A critique initiated by the plant mWger revealed that just prior to the event l&C technicians identified several discrepancies between the new and old relay and sought guidance from l&C supvvisio However, due to miscommunication between the l&C supervisor and the system engineer, the I&C technicians were informed that the replacement relay was acceptable. BECo initiated PR 97.9821 to document and evaluate this event including an apparent cause analysi The inspector recognized tne positive action of the I&C technicians to stop work and seek clarification when the replacement relay serial number did not completely match the one documented in the work package. However. l&C supervision again missed an opportunity to correct a problem at an early stage. The inspector discussed the effectiveless of I&C supervision with the plant manager who indicated that efforts were underway to provide additional training and clarification of his expectations for l&C supervisor The inspector reviewed the apparant cause analysis and interviewed several personnel directly involved in the event. The original work control planner placed three 24 vde raiays on reservation (i.e., available for use), which were in stock, and realized that a fourth relay with a 125 vde rating was not in stock and needed to be ordered. Tha original planner

) placed the work package on a " parts hold" status pending the arrival of the fourth rela Subsequently, the original planner was not available for an extended period of time and a

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l second work planner and an l&C supervisor reviewed the package and changed the status from " parts hold" to " task ready". The second planner and l&C supervisor did not realize that a fourth relay of a 125 vde rating was required and had been ordered. This action violated procedure 1.5.20, Work Control Process, section 7.5[3), Task Ready Review, which specifies that parts are reserved / withdrawn for use prior to designating the package as task ready. The second planner and l&C supervisor thought that only three relays were needed and did not realize that a fourth relay was on order and had not arrive Additionally, the l&C supervisor violated step 7.5[4] of the Work Control Process, which reouires a hands-on-parts verification for each job and that parts should be pre staged one week prior to the week that the work is planned. Collectively, these two failures to adhere with procedures are considered a violation of Technical Specification 6.8. (VIO 9713 01)

The inspector also learnod that the work control group uses an !nformal process of including any stock material authorization forma (SMA), otherwise known as material request forms,in the walk control package. In this case, the first planner reserved the three 24 vde relays and ordered the 125 vde reby, but did not include a copy of the SMA

, in the work package. This contributed to the inadequate turnover between the original and second planner since the original planner knew that a fourth relay of 125 vde rating was

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required. The licensee informed the inspector that the informal practice of including SMAs in the work package would be reviewed and enhanced, as needed. Other contributing factors to this event involved the work package controls which did not clearly highlight the difference in the relay types. Additionally, an engineering document which specified the relay details included many different types of applications and was confut!ng. In summary, this event resulted from multiple human performance issues which primarily involved work control planners and l&C personnel, Conclusions Routine and emergent maintenance tasks were completed by experienced workers with gene ally good results. A planned emergency diesel generator LCO maintenance outage

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was better planned and executed than one during the previous period.

l The NRC identit ed two l&C human performance issues with the potential use of expired thread sealant and the inadvertent opening of a petcock valve on the lube oil strainer for the "A" EDG, The ATWS system was declared inoperable when l&C technicians inadvertently installed a 24 vde relay into a 125 vde application which was also considered a human performance 4 issue. The relay overheated when placed in service, disclosing the human error during the activity. The event was caused by a work planner and an l&C supervisor, who changed the status of the work package from "on hold for parts" to " task ready," when, in fact, all parts were not available. Technicians in the field identified discrepancies with the replacement relay, but were incorrectly informed by the supervisor and a system engineer that the relay was acceptable. These errors resulted in a violation of the work control process requirement . - - - _ . _ . . - - . - - _ - - - - _ -

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in three human performance related instances during this period, l&C supervisors missed opportunities to correct adverse conditions, and on one occasion this led to inoperable

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important to-safety equipmen M2 - Maintenance and Material Condition of Facilities and Equipn ent M2.1 (Closed) LER 50 293/97 25; Shutdown due to Inoneratile Main Steam Isolation

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Valve (MSIV)

$

. Insoection Scone (62707)-

o l On November 22,1997, MSIVs AO 2031C and 2B did not indicate full closed djring the ,

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push button portion of procedure 8.7.4.4, TMain Steam isolation Valve Operability,~ 60_ i F Percent Power." The MSIVs were declared inoperable and the plant commenced a technical specification required shutdown. The inspector reviewed the actions taken to address the !

slow closure failure of the MSIV's.

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- Observations and Findinas

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. The event occurred at 50 percent power during the performance of a technical specification (TS) surveillance to check the containment isolation function. The MSIVs

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employ a pneumatic cylinder operator and closing spring as separate, locally stored enerr;y "

sources for rapid closure. Both the spring force and pneumatic energy are capable of independently closing the valves against full reactor pressure as indicated in Updated Final Safety Analysis Report (UFSAR) section 4.6. During the slow closure test air is slowly bled from the air cylinder bottom through a test valve to test the capability of the spring to -

3 close the valve. MSlVs 1C and 2B both passed the fast closure test within the required time of 3 to 5 seconds, but failed the slow closure test.' The remaining 6 MSIVs passed -

l lboth the fast and slow closure test ;

$ PNPS TS require that the MSIVs be exercised by partial closure and subsequent reopening i -twice weekly in addition to being tested in accordance with the in-service code testing (IST) requirements. The IST program testing program cold shutdown justification CS-17 requires partial stroking of the MSIVs quarterly and full fast closure testing durng cold

- shutdown. BECo performs the full closure test quarterly vice during cold shutdown to

.._ Gnsure freedom of movement of the main poppet and to provided additional exercise of the ' four way control valve. The slow closure test is an additional (non safety) UFSAR design l- requirement. The inspector determined that the performance of the quarterly slow closure

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test was a conservative measure by BECo to detect MSIV degradation at an early stag BECo investigation revealed that the jailure of the two MSIVs to fully close was due to the relaxation of the closing springs (approximately_23 percent degradation). The MSIV vendor manual did not provide any specific criteria for the frequency of actuator overhauls or

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spring life. A contributing cause was increased friction between the spring plate and .

9 actuator stanchions due to improper adjustment or loosening of the cam rollers. Corrective actions taken included replacement of the closing springs for AO 203-1C,2B, and 2D. In i- addition, valves AO-203-1C and 2B valve actuators were overhauled, live load packing was

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installed to reduce friction, and the cam rollers were checked and adjusted as necessar The remaining MSIVs had live load packing or were repacked with live load packing, and the cam rollers were inspected, lubricated and adjusted as necessary to restore any margin I. st caused by spring relaxation. BECo plans on revising the MSIV maintenance and

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surveillance procedures to time the slow closure test to trend and better detect valve degradation. The actustor springs for the remaining five MSIVs are scheduled to be replaced during the next refueling outag The inspector noted that BECo has determined that this condition constituted a functional failure as defined by the maintenance rule. A main steam system functional failure is any failure that causes a forced shut dowr: or power reduction of greater than eight hours. The main steam system has not experienced any other functional failure within the past two years, therefore the system was not placed in a maintenance rule (a)(1) status, Conclusions An unplanned shutdown resulted from degraded MSIV actuator closing springs which was considered a maintenance functional failure. BECo expanded the corrective actions to restore margin on all MSIVs. The use of a quarterly surveillance test, beyond the TS rnquirements, to detect actuator degradation at an early stage is a positive initiativ M4 Maintenance Staff Knowledge and Performance M4.1 Feedwater Reaulatina Valve Maintenance troubleshootino jngnpction Scooe (62707.92700)

The inspector reviewed the results of the maintenance troubleshooting efforts on the "A" FRV (i.e., FV 642A) which failed open and caused the automatic scram on December 6, 1997. The operational aspects of this event are discussed in Section 02.1 of this repor The inspector visually examined the air positioner, interviewed personnel and reviewed PR 97.976 Observations and Findinos Instrument and Control (l&C) technicians found that a pilot valve clip, located in the Bailey valve positioner for FV-642A, was found displaced. This caused the pilot valve stem to drop, bleed off air and causing FV-642A to go full open. After restoring the valve clip to the normal configuration, FV 642A functioned properly. As a precautionary meestre, BECo installed a whole new valve positioner. The BECo investigation team determined that the most probable cause of the valve clip becoming displaced was during some recent corrective maintenance conducted during the November MSIV shutdown cutage. At that time, l&C technicians performed work inside the Bailey air positioner and probably inadvertently bumped the valve clip from it's normal conditio During the maintenance troubleshooting on FV-641A, the inspector observed excellent teamwork between I&C and engineering personnel. BECo initiated a formal root cause I

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11 I analysis for PR 97.9765 which had not been completed at the end of this inspection period. The inspector noted that the scram on December 6,1997 resulted from a maintenance preventable functional failure of the feed water system. Based on this failure, the system engineer classified the feed water system as an A.1 system under the maintenance rule criteria. This criteria allows only one failure every two years and a previous system failure occurred in February 1997. The inspector determined that BECo properly applied the maintenance rule criteri in addition to bumping the valve clip, the inspector noted that on two other occasions .

during this inspection period, equipment was inadvertently altered or damaged during the conduct of work activities. For example, the inspector identified during the "A" EDG .

outage that l&C technicians erroneously bumped open the vent valve on the lube oil

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strainer. Further, BECo identified that during MSIV corrective maintenance in the steam tunnel, an MSIV accumulator drain valve stem was inadvertently stepped on and damaged l which required corrective maintenance. An independent post-scram BECo team assessed these eve *2 and determined that increased worker and supervisor sensitivity was needed to prevent recurrence of " collateral damage". Conclusions -

Maintenance troubleshooting efforts were effective in identifying the most probable cause why FV 642A failed open on December 6,1997 resulting in a reactor scram. The BECo post trip review team determined that an l&C corrective maintenance activity during the MSIV shutdown outage inadvertently interfered with the proper positioning of the valve clip located in the valve positioner. The system engineer properly applied the maintenance rule criteria for this maintenance preventable failure. Corrective actions were thoroug M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) Violation 50-293/97-01-02:Inadeouate corrective actions for three poerator work-arounds and also for a missed standbv switchvard batterv surveillanc The NRC identified two examples where corrective actions were not timely and effectiv The first example involved the adequacy of corrective actions for a licensee identified missed back up switchyard battery surveillance. The corrective action of changing the related master surveillance tracking program (MSTP) node did not prevent possible recurrence, in response to the violation, BECo changed the due date of the missed surveillance in the MSTP to prevant possible recurrenc The second example of inadequate corrective actions involved three operator work around conditions that adversely impacted operator response during low power post scram conditions on February 15,1997. Two of the three work arounds (e.g., control rod full-in lights and the reactor water cleanup (RWCU) letdown valve CV 1239 operation) were longstanding and compensatory measures had been incorporated into various procedures, in the response letter to the violation dated April 17,1997, BECo admitted to the violatio Corrective actions were taken and planned to address each of the three specific degraded equipment condition l

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Main steam drain line isolation valve MO 220 3 was replaced with a new valve in refueling outage no.11 (RFO11). This eliminated the work around condition of dispatching an operator to manually operate the valve which is not readily accessible. New valve trim was installed in CV 1239 and the air operator was calibrated. This corrective maintenance eliminated the work around condition of installing a mechanical gagging device and allowed finer control of letdown to help prevent inadvertent RWCU isolations due to erratic letdown flow through CV-123 The last work around condition irivolved control rod full in lights that sometimes do not illuminate after a scram causing the operators o enter EOP-2, Failure to Scram. The generic aspect of this BWR pr(Clem is discussed in General Electric SIL 532," Full-In Control Rod Position Indication," dated March 27,1991. SIL 532 describes possible thermal effects on control rod drive magnets with the potential resultant loss of all rods full-in indication ufter a scram. Experience at PNPS has revealed that allowing rods to settle out from the over travel condiuon will usually cause the full-in lights to illuminate. BECo engineering personnel prepared plant design change (PDC)97-10A, Control Rod Position Upgrade - Phase I, which was implemented as an interim measure in August 1997. The design change modified each of the 145 control rod probe buffer cards to allow the process computer to read either the full-in or full-in over travel reed switches. In the response letter to the violation, BECo committed to develop and implemerst a final design change by the end of the first quarter in 1998 providing the work could be performed on-lin During the reactor scram on December 6,1997, six control rods did not indicate full-i When operators closed valve 30125, CRD Charging Water Supply Valve, in accordance with step 5.0(2](c) of procedure 2.1.6, " Reactor Scram," to maintain reactor vessel level control, the six control rods " settled out" form the over travel condition and provided full-in indication. Although the interim modification did not fully resolve the temporary loss of full-in indication following a scram, the change allowed engineering to retrieve more data to aid in development of a final sc!ution. At the end of this inspection period, engineering personnel were considering a modification to add a motor actuator to valve 301-25 to allow control room operators to quickly cl .e the valve following a scra The inspector reviewed proceJure 2.1.6 relative to guidance provided for resolving control rods which do not indicate full-in after a scram. Section 5.0, Subsequent Operator Actions, cortained a note to reset the scram to allow the control rods to settle ou However, th3 note specifies that the scram can only be reset after all scram conditions clear which requires additional time after the scram. Also, in section 2.0, Discussion, includes guidance to ascertain control rod position, but the inspector noted that the guidance did not include shutting the 301-25 valve. Procedure 2.1.6 referenced closure of 30125 for reactor vessel water level control purposes onl The inspector discussed the need to clarify procedure 2.1.6 with the assistant operations department manager (ODM). The assistant ODM initiated a review of procedure 2.1.6 and subsequently operations support personnel issued revision 40 to procedure 2.1.6, dated December 31,1997. Revision 40 added guidance to close valve 301-25 after a scram to

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allow the control rods to resettle. The inspector had no further concerns with the procedural guidance in this regar BECo took progressive steps to resolve the temporary loss of full-in indication following a scram. A long term modification remained under evaluation inc!ading the installation of a motor actuator on valve 30125. Other long term solutions were alsc being considere Procedure 2.1.6 did not .nrovide clear guidance to close valve 30125 to allow the control rods to settie out; however, a procedure change was made during this period to provide more guidance. BECo committed in the response to violation 97 01-02 to implement a long term solution to resolve the temporary lo s of the full in lights, in addition to the specific corrective actions for each of the three aforementioned degraded conditions, BECo formed an operator work around multi-disciplined team to evaluate the need for broader corrective actict:s and to determine the full problem extent. BECo reviewed 87 operations procedures to identify potential compensatory actions contained in operation department proceoures to work around degraded conditions. A few other degraded conditions were identified and added to the operator work around list. Also, the team expanded the definition of what constitutes a work around condition. A training module was prepared for both engineering and operations personnel to provide increased guidance for the identification and resolution of work around condition Based on the corrective actions taken and planned by BECo, the inspector determined that violation 97-01-02 is close M8.2 (Closed) URI 50-293/96-80-02: Work Control Issun Two licensee identified work control issues were identified in early 1996 following a substantial change made to the work control process. Both events involved issues with post work tests (PVVT) due to weak work control packages. One issue related to a missed RHR system PWT which was documented and treated as a non-cited violation in NRC inspection report no. 50-293/95-26. The second PWT i.esue related to a missed visual excmination. The inspector reviewed the specific corrective actions taken to address the missed visualinspection and found the actions were timely and comprehensive. This licensee identified and corrected violation is being treated as a Non-Cited Violation (NCV 97-13-10), consistent with Section Vll.B.1 of the NRC Enforcement Policy. These two issues collectively indicated a weakness in the new work control proces As a result, BECo changed the new work control process to better define responsibilities of the members of the new maintenance team approach. The inspector attended training sessions held by the work control department manager with work control personnel. The training session was detailed and pertinent examples were used to emphasize salient points, inspector review of the two events identified no violations of technical specification operability requirements. Generally good performance was noted by the inspector in the execution of PWTs during the previous twelve months indicating that the corrective actions were effective. Based on this review and the corrective actions taken, the inspector considers this unresolved item is close _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

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Ill. ENGINEERING E1 Conduct of Engineering E (Ocen) VIO 50-293/9713-02: Reactor Vessel Flanae Metal Temocrature Indication Jnsoection Scoce (37551)

On December 2,1997, during the sctor heatup, the indlcated reactor vessel flange to adjacent vessel shell temperature differential exceeded the technical specification (TS) limit of 145'F. Subsequently, the licensee determined that no violation of the differentia l temperature existed. BECo questioned the adequacy of the indicated vessel /shell temperature differential readings due to problerns previously experienced with the temperature elements. The temperature instruments used to monitor the tempe,z.ture differential were not the normally installed temperature elements. A temporary modification was made during the cycle 11 refueling outage (RFO-11) due to problems with the original instruments not tracking properly. At the time, BECn ordered replacement parts to be installed during RFO12. The inspector reviewed the conditions leading to the -

event (Section 01.1), the maintenance history of the instrumentation, and the temporary modificatio . Observations and Findinas There are three non safety related teTiperature elements used to monitor vessel flange temperature which are located approximately every 120 degrees around the reactor vessel fienge. Prob! oms were experienced with the normally installed temperature recorders during the end of hFO-11. Temporary modification (TM) 97 29 was implemented to provide reactor vessel flange metal temperature for plant operation during operating cycle 12. The original thermocouples could not be repaired without removing the drywell head and insulation package, and replacement parts were not available. Three temporary magnetic mounted thermocouples were installed on tFe vessel flange at two locations approximetely 180 degrees from one another, in a.Jdition, new extension wires were l installed from the thermocouples down to the existing plant junction box to monitor reactor vessel temperatur The inspector reviewed TM 97 29 and noted that the thermocouples used were the same type as the degraded ones, but were attached by a magnetic probe vice installed in a holder. The magnets used for the TM were rated for ten pound pull. BECo concluded that the TM was functionally equivalent to the original design and thus no unreviewed safety question existe The inspector reviewed the maintenance records for the vessel flange temperature elements, since installation of TM 97-29, and identified that several problems have been experienced. On May 30,1997, point 3 (TR 263-105) temperature indicated over 100*F lower than point 2 (TR-263-104). A problem report was writton to document the condition. On November 26,1997, during the forced shutdown to repair the main steam isolation valves, point 3 failed down scale. Inves.igation by maintenance personnel revealed that the temperature element for point 3 was found separated from the vessel flange inspection of the other elements revealed that point 2 had also separated from the vessel flange. Additionally, the terminal screws for these two temperature elements were

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loose. The inspector's review of problem reports (prs) revealed that no PR was written to document that the temperature elements had separated from the reactor vessel flang Discussions with the maintenance personnel who identified this condition revealed that they had informed a system engineer of the finding BECo's ir.vestigation into the December 2,1997, failure of the temperature elements identified that point 1 (EPIC) and point 2 were incorrectly positioned. The temperature elements were mounted on the refueling bellows skirt, which attaches half way down on the vessel flange. Point 2 was also found to have a bad connection. Point 3 was correctly positioned on the vessel flange, but was found to have a broken spring that is used to hold the thermocouple in place against the vessel flange. When the temperature recorder for point 2 was positioned back onto the vessel flange, the indicated temperature increased approximately 50'F. This was consistent with a reading taken wi+h a hand held pyrometer. Point 2 it used to verify that the vessel f!ange and shell temperatures do not differ by more than 145'F during reactor heatu Corrective actions for the December 2,1997, event includec using a special placement tool to position the temperature elements in the proper positions, replacing point 3 with a button type magnet, and verifying that the temperature elements were reading properly by comparing them to temperatures obtained with the hand hold monitors. BECo determined that the temperature elements had been improperly oositioned by the maintenance craft i during previous troubleshootin During the December 7,1997, startup following the automatic reactor scram (Section O2.1) point 2 again was not operating properly. Inspection by BECo revealed that point 1 and point 2 were not making contact. Point 2 was also found to have a bad connecto BECo replaced the electrical connectors with a ceramic type that is rated for higher temparatures, ciecaed off the vessel flange and reinstalled the temperature detector The corrective actionc taken for the temporary temperature elementu separating from the reactor vessel fisnge were not effective in preventing repeat occurances. The inspector noted that BECo's investigation into the cause of the temperature elements being improperly positioned on December 2 and December 7 did not take into account the November 26,1997 event. Procedure 1.3.121, " Problem Report Program," revision 3, requires that hardware and non-haidware related prob!ams be documented on a PR. This includes failures, malfunctions, deficiencies, human errors, defective or degraded material or equipment, and non-conformances. For identified t,quipment problems a direct cause analysis or root cause analysis is performed to determine the cause and develop corrective actions. The failure to document hardware problems on a PR (the November 26 temporary temperature elements separation from the reactor vesses flange) and take effective correctivo actions on three separate occasions for the temperature elements dislodging from the vessel flanno is a violation 50-293/9713-02of Appendix B, Criterion XVI for inadequate corrective actions and BECo procedure 1.3.12 & ,

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The inspector questioned BECo regarding the adequacy of the TM and safety evaluation based on the movel.ient/ separation of the temporary temperature elements on three different occasions. BECo was evaluating the adequacy of the TM and determining the cause of the temperature element movement at the end of the inspection period, .GDDplusions Operators were challenged during plant heetup and cooldowns due to reliability problems from a temporary modification made on the reactor vessel flange temperature element BECo did not adequately evaluate the cause of and develop effective corrective actions to preclude the recurrent instances of the temporary temperature detectors from becoming dislodged from the reactor vessel flange. The elements moved or were separated from the reactor vessel flange on three separate occasions. The failure to adequately evaluate and correct the problem is considered a violatio E1.2 (Open) (IFl 50-293/97-13-03/04/05): Generic Letter 8910 Motor-Operated wag Proaram Review l Insoection Scoos (Tl 2515/109)

On June 28,1989, the NRC issued Generic Letter (GL) 89-10, " Safety-Related Motor- .

Operated Valve Testing and Surveillance," which requested licensees to establish a :

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program to ensure that switch settings for safety-related motor operated valves (MOVs)

were selected, set, and maintained properly. Seven supplemen:s to the GL were issued to i provide additionalinformation and guidance on development of programs NRC inspections at Pilgrim Nuclear Power Station were conducted based on guidance contained in NR Temporary Instruction (TI) 2515/109," Inspection Requirements for Generic Letter 89-10."

The results of previous MOV program inspections were documented in Inspection Reports 50-293/92 80and 50-293/93 2 ;

The purpose of this inspection was to examine the actions implemented at Pilgrim, and to determine if those actions were sufficient to warrant closure of the NRC staff's review of the GL 89-10 MOV program. Documents and calculations were reviewed for the following ;

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MO 1201-02 Reactor water cleanup inboard suction isolation MC 120105 Reactor water cleanup outboard suction isolation MO 1301-16 Reactor core isolation cooling inboard steam supply isolation MO 1301-17 Reactor core isolation cooling outboard steam supply isolation MG 2301-04- Hi@ pressure coolant injection inboard steam supply isolation MO 2301-05 Hign pressure coolant injection outboard steam supply isolation MO 4085B Reactor building closed cooling water header outlet isolation

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17 Observations and Findinas General BECo's MCV program was documented in Nuclear Organization Proceduto (NOP) 92M1,

" Motor Operated Valve Program," usted Septembr 30,1997. Valve factor justifications were included in individual thrust calculations. Calculations M 767, " Basis for Rate of Loading Assumptions," dated November 8,1907, and M 768, " Bas,is for Lineer Extrapolation Assumptions," dated November 11,1997, contained the licensee's justifications for load sensitive behavior and extrapolation of dynamic test results to design basis condition Grouoina and Valve Factors BECo did not establish valve groups in the Pilgrim MOV program. Procedure NOP92M1 allowed the application of test data from valves that were equivalent in design, functional characteristics, and operating condition (e.g., fluid temperature, subcooling, differential pressure). The data were then used to establish th thrust requirements for valves that could not be tested under dynamic conditions. Rather than utilizing groups, the licensee applied In plant and/or industry test information on a valve by valve basis. The inspectors independently evaluated the valve population by dividing the valves into groups based on valve manufacturer, type, size, and ANSI pressure class. This allowed the inspectors to assess the methods used to justify applied valve factor Concerning valve types for which l'ttle or no in-plant test results were available, BECo assumed generic valve factors anci referenced friction coefficients obtained from the Electric Power Research Institute's (EPRl's) prototype test program that was performed in support of the Performance Prediction Methadology (PPM). The valve f actors used in BECo's design calculations were more conservative (i.e. conside. ably higher) than suggested by the EPRI data alone, in that sense, DECO did not rely solely on the EPRI dat In Section B uf the NRC's Safety Evaluation of EPRI Topical Report TR 103237,"EPRI Motre Operated Valve Performance Prediction Program," dated March 15,1996, the NRC established its position in this regard, indicating that there's not sufficient evidence to justify rep lying those friction coefficients directly in industry equations to predict valve thrust reouirements as valve factors or friction coefficients. Therefore, BECo's selective use of EPRl's friction coefficients was not acceptable whhout further technical justificatio However, based on BECo's valve f actor assumptions and current control switch settings, the inspectors did not identify any immediate operabiliry concern The following non-tested valve types were identified as requiring additional valve factor information to support current program assumptions:

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i Valve Manufacturer ANSI Applied Valve Number of '

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Size Class Factors Affected '

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Valves j Close -Open !

Flex Wedge Gate Valves i 16" Anchor /Da. ling 150# 0.50 0.50 11 t

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18"

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10" Anchor / Darling 900# 0.50 0.60 1 4" Powell 150# 0.50 0.50 2 l 20" Powell 600# 0.60 0.60 1 4" Powell 900# 0.70 0.70 1 -

10" Velan 300# 0.60 0.60 4 8" Velan 600# 0.70 0.70 1 3" Velan 600# 0.50 0.60 2 20" Wal worth 600# 0.60 0.00 1 3" Westinghouse 1525# 0.55 0.55 2

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3035#

Double Disc Gate Valves 24" Anchor / Darling 900# 0.75 0.75 2 Globe Valves 4" Anchor / 1500# 1.10 1.10 1 l

Darling

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2" Powell 600# 1.40 1.40 1 4" Powell 900# 1.10 1.10 1 6" Velan 300# 1.10 1.10 2 10" Velan 900# 1.10 1.10 1 18" Walworth 600# 1.10 1.10 2

, Additional information can be obtained from further in plant dynamic testing, the PPM (as applicable), or other applicable industry sources. ln a latter dated December 11,1997, BECo committed to obtain additionalindustry information or utilize the EPHI PPM to bolster current valve f actor assumptions. BECo expects to complete this effort by June 30,199 This item remains open pending review of the data. (IFl 50 293/97 13 03)

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kgirculation Pumo isolation Valves in design basis thrust calculations for the recirculation pump suction and discharge isolation valves, the licensee assumed three different valve factors. The valves are 24-mch,900 psi-Class, parallel disk gate valves manuf actured by Anchor Darling Compcn Per calculation M 503,"AC MOV Design Basis Review," the safety function of pump discharge >alves MO 202 6A/ Bis to close per low pressure coolant injection loop selection logic to din et emergency core cooling flow to the intact recirculation loop during a loss of coolant accident (l.OCA). The calculated differential pressure for this event is 32 psid, and in calculation M 530, "MOV Thrust / Torque Calculation for Priority 3 MOVs," a bounding valve factor of 0.75 is assumed, in accordance with Section 4.3 of the Pilgrim Updated Final Safety Analysis Report (UFSAR) and operating procedure 2.4.22," Failure of Recirculation Pump Seal," the pump suction and discharge valves also are used to isolate a recirculation pump in the event of a cetastrophic shaft seal failure. For valves MO 202 5A/B,the differential pressure assumed for this event is 200 psid, based on General Electric Company purchase specification 21 A1084 and NEDC 31871,"BWROG Report on the Operational Design Basis of Selected Safety Related MOVs in Response to Generic Letter 8910," dated April 1991. For a seal failure event, the licensee assumed a valve factor of The inspectors considered the licensee's LOCA valve factor assumption (0.75) to be high based on the results of in situ and EPRI performance prediction program test results for similar valves. Ilowever, the assumptions of 0.5 and 0.4 for were not technically justifie Also, the 200 psid differential pressure assumed for the pump seal failure event appeared to be a generic industry value, with no technical basis specific to Pilgrim. Because all four valves are identical, but three different valve f actors are assumed, its ur, clear which design basis scenario is more limiting (namely, either the LOCA or seal failure), particularly since the 200 psid assumption is probably too high. While the design basis capability of the '

pump discharge valves was apparently adequate (based upon current switch settings), the correct operating conditions (and valve factor assumptions) for the pump isolation valves (MO 202 4A/B) required further clarification. This matter will be tracked as an inspector followup item. (IFl 50 293/97 13 04)

Load Sensitive Behavior The inspectors considered BECo's load sensitive behavior margins to be acceptable for GL 8910 program closure. BECo's load sensitive behavior study documented an analysis of 33 in plant gate and globe valve data points. Using this analysis, the licensee selected a 4% bias margin based on the mean of the data, and 11% as the random margin that was based on 2 standard deviations of the data. The random value was combined with other random uncertainties using the square-root sum of the squares methodology. The results for valve MO-40098 ( 22.9%) were not included due to the large negative valee (negative values indicate that more thrust was available at torque switch trip under dynamic a

cond;tions).

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Stem Friction Coeffielent The licensee assumed a stem friction coefficient of 0.15 for non dif ferential pressure tested valves. However, at the beginning of the inspection, BECo had not completed a stem friction coefficient study for Pilgrim. The licensee had obtained a large amount of stem friction coefficient data as part of its MOV test program. These data included results obtained under static and dynamic conditions in the close and open directions. The setup method, which included a 15% load sensitive behavior margin applied to the open thrust requirements and a 3% margin for lubricant degradation, was intended to account for any higher than cxpected stem friction coefficient conditions. Available margins were then determined based on the highest stem friction coefficient value obtained from dynamic testin During the inspection, the licensee provided draft calculation M 772, " Evaluation of MOV Coefficient of Friction," to support the current assumptions. The analysis used the test results from 68 statically tested MOVs and 31 dynamically tested MOVs. Each coefficient of friction was calculated based on an average of three to five tests, and upper confidence limits were calculated based on a 97.5% Students T curve, inspectors independently analyzed the test data in the draft calculation and aoncluded that a generic (Pilgrim GL 89-10 program assumption) stem friction coefficient of 0.20 was justified. Design basis thrust calculations for certain valves that were not diagnostically tested will need to be revised from the current assumption of 0.15. The inspectors determined that the resulting reduction in the capability of these valves was not significant. The NRC will review calculation M 772 after it has been verified and approved by the licensee. (IFl 50 293/97 13 05)

Extracolation of Dynamic Test Data The licensen's justification for linear extrapolation of dynamic test data did not include EPRl's latest recommendations for identifying the disk loads that are necessary to ensure that test results are reliable. The licensee Intended to update its program by a review existing test conditions to ensure that adequate uisk loading was obtained, Conclusions As discussed below, the licensee demonstrated the design-basis capability of its GL 8910 MOVs. However, the licensee's valve factor assumptions for some non differential-pressure tested MOVs need to be strengthened. BECo committed to obtain additional information to bolster the assumptions by June 30,1998. The licensee appropriately adjusted MOV program design assumptions in response to industry information. Program assumptions regarding load sensitive behavior, stem friction coefficient, and extrapolation of test data were acceptable for GL 8910 program closur . ~ . - - . . . .

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E1.3 Motor Onerated Valve Desian Basis Canability insocction Scone The inspectors reviewed design basis thrust calculations, capability assessment packages, and tett reports for the selected MOVs. The purpose of the review was to assess DECO's efforts to establish design-basis capability for all MOVs in the GL 8910 program at Pilgrim, Observations and Findinas The inspectors reviewed bummary documents that identified the available thrust margins for all MOVs in the program and identified several valves that had negative design margins based on program assumptions. On November 9,1997, the licensee issued Problem Report (PR) 97 9684 to address the negative margins for the valves listed in the following table:

Valve Original Design Revised Margin Margin MO 130122 4% Close 68 %

MO 4002 13% Close 29 %

MO 40098 16% Close 9%

MO 4085A -15% Close 37 %

MO 40858 20% Close 17 %

MO 4060B 11% Open 5%

Except for valve MO 130122, all of the MOVs were tested under dynamic condition BECo calculated revised margins by replacing the design packing loads with no.v values

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that bounded the actual packing loads measured during diagnostic testing. The licensee also determined that the design basis differential pressure for valve MO 40858 was too high based on pump shutoff head conditions that would not exist when the valve is operated. Based on the information contained in PR 97 9684,the inspectors agreed with

the licensee's operability assessment for the valve The inspectors also reviewed the available thrust margins for the remaining valves in the MOV program and determined that the following valves have a relatively small amount l (between 5% and 10%) of margin in their safet/ function stroke directio MO 2301-33 MO 1301-49 MO 230104 MO 4009B MO 120180 MO 130153 MO 2301-10 MO 4060B MO 3813 At the time of the inspection, the licensee had not fully evaluated the need to improve the margins of these valves. The issue of " margins" is acceptabis for GL8910 program closure, DECO indicated that it intends to inform the NRC of its plans regarding the valves by June 30,199 . 1

22 BECo's MOV program included eight Pratt symmetrical butterfly valves; six valves were located in the Salt Service Water (SSW) system and two valves were located in the Reactor Building Closed Cooling Water (RBCCW) system. The licensee performed diagnostic dynamic tests on all of the butterfly valves. The SSW system valve tests showed that the vendor's torque requirements were higher than necessary; however, the licensee was unable to obtain differential pressure measurements during the RBCCW valve tests. Therefore, the licensee was unable to extrapolate the test results to design basis conditions. Further, a downstream temperature control valve (closed during the dynamic tests) affected the tests' differential pressure condition Based on static tests of the RBCCW valves, the licensee determined that valve MO 4083 had a 29% margin and valve MO 4084 had 57% margin. The inspectors noted that the total seating torque measured during the dynamic test of valve MO 4083 was 269 f t l This value closely approximated the vendor's prediction of 283 ft lb. While the inspec*. ors had no concerns regarding the operability of the RBCCW valves, more information will be needed to validate the vendor's torque requirements for these two butterfly valves as part of BECo's long term MOV progra BECo has begun to implement an in-place MOV diagnostic system (i.e., the Sentry system developed by Teledyne) that automatically gathers diagnostic data each time that an MOV is stroked. This innovative and useful system is permanently installed and gathers direct force measurements for later analysis. The licensee has installed systems for six MOVs and plans to install approximately 60 more systems in the future to support the periodic verification program. Given the system's ability to gather diagnostic data every time a valve is stroked, the inspectors considered the implementation of this system to be a strength, Conclusions Based on available margins, the licensee demonstrated design basis capability for each of the GL 8910 MOVs at Pilgri E1.4 Deslan Calculations Insoection Scong The inspectors reviewed the following MOV design calculations to verify tnat calculations were updated to reflect the results of diagnostic testing and other plant changes:

M-553, " Maximum Pressure Differential for DC MOVs" M 563, "AC MOV Design Basis Review" PS 132(133,134,135)," Electrical Performance and Stroke Timing Evaluation of Priority 2(3,4,5) AC MOVs"

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23 Observations and Findinag The inspectors reviewed engineering evaluations performed por instruction NEDWI 449,

" Dynamic Test Evaluation." The inspectors verified that, in most cases, the assumptions used in MOV design calculations were updated to reflect diagnostic test results. However, three relatively minor instances were identified in which design calculations had not yet been revise The design basis differential pressure for reactor building component cooling water system isolation valves MO 4002 and MO 4009 was 58.9 psid in calculation M 563. However,in the diagnostic test evaluations of valve MO 4002, the licensee extrapolated the test results to 54 psid. The lower differential pressure also was incorporated into thrust calculation , "MOV Thrust / Torque Calculation for Priority 4 Valves," Revision 7, dated November 8,1997. The evaluation of valve MO 4009 extrapolated the dynamic test data to 67 psid vice 58.9 psid, and calculation M 530, *MOV Thrust / Torque Calculation for Priority 3 MOVs," Revision 6, dated November 8,1997. (A note in this calculation indicated that calculation M 563 needed to be updated for the new differential pressure.)

Calculation M 553 contained two different design basis differential pressures for residual heat removal shutdown cooling isolation valves MO 100150 and MO 100147, based on maximum reactor pressures of 110 psig and 80 psig. Plant design change 9412,

" Shutdown Cooling Pressure Switches Setpoint Change," dated May 19,1994, reduced the analytical differential pressure limit, providing additional operational margin for the valves. Calculation M 553 needs to be revised to delete the differential pressure based on a reactor pressure of 110 psi Calculation PS 79, " Emergency Diesel Generator Loading," dated January 19,1996, calculated electricalloads for several MOVs that did not reflect current field conditions, in most cases, however, the calculated loads were greater than actualloads, and the calculation's conclusions remained conservative. The licensee was aware of the I inconsistencies and had qualitatively evaluated the effects of MOV modifications in several Calculation Comment Sheets. The licensee also informed the inspectors that a major revision of the dieselloading calculation was in progres Collectively, these instances represented inconsistencies between design calculations and field conditions. Failure to update the design calculations was a violation of 10CFR50 Appendix B, Criterion lit which requires the design basis of safety related components to be translated correctly into specif; cations, drawings and procedures. Neither the conclusions of the calculations, nor the operability of the MOVs were affected adversel Therefore, this f ailure constitutes a violation of minor significance and is being treated as a NCV consistent with Section IV of the Enforcement Policy (NCV 50 293/9713 06) Conclusions Overall, design-basis MOV calculations were updated to reflect the latest technical information. Isolated cases were identified in which calculations needed to be updated, but the licensee was aware of the discrepancies and revisions either were planned or in progres a

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E2 Engineering Support of Facilities and Equipment E Trackina and Trendina Proaram Insoection Scone The inspectors reviewed procedures NOP92M1, " Motor Operated Valve Program,"

NESDWl 453, *MOV Plant Performance Monitoring," and the latest two year cycle MOV

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report to assess BECo's program for tracking and trending MOV failures and performanc Observations and Findinas Procedure NOP92M1 defines the responsibilities of the Nuclear Organization for all aspects of the GL 8910 program. The tracking and trending program is assigned to the MOV system engineer, who trends MOV performance in accordance with work Instruction NESDWl 453. Status reports consisting of maintenance summaries, diagnostic test results and corrective actions were generated after each refueling outage. The items tracked by the program were consistent with those contained in Attachment A (Summary of Common Motor Operated Valve Deficiencies) of Generic Letter 8910. The inspectors reviewed the latest status report and noted no recurring MOV performance problem BECo also implemented administrative controls adequate to ensure that industry information, such as 10 CFR Part 21 notices and NRC Information Notices, were routed to the responsible staff for review and action. For example, the licensee responded appropriately to potentially adverse industry information regard!ng MOV performance contained in NRC Information Notice 96 48," Motor Operated Valve Performance issues,"

dated August 21,1996, and other industry sources. BECo's methods of calculating Limitorque valve actuator capability included the use of: (1) pullout efficiencies for both stroke directions, (2) a 0.9 application factor for alternating current motors, and (3) an exponent of 2.2 to calculate degraded voltage factors for alternating current motors, Conclusions BECo's tracking and trending program was consistent with the guidelines established in Generic Letter 8910. The licensee made effective use of industry informatio E2.2 Periodic Verification of MOV Performance Jnsoection Scope The inspectors reviewed Exhibit 1 of Procedure NOP92M1, Motor Operated Valve Program," which contains a description of BECo's periodic verification program, Observations and Findinas in its response to Generic Letter 96 05," Periodic Verification of Design Basis Capabi!!ty of Safety Related Motor Operated Valves," BECo committed to develop procedures and to implement a periodic verification program by December 31,1997. At the time of the inspection, the procedures were not yet writte _ _ _

O o

The licensee is a participant in the Joint Owners Group (JOG) program for MOV periodic verification. The program consists of a mix of static and dynamic diagnostic tests conducted at a frequency determined through a consideration of MOV design capability and risk significance. Each of the 85 active MOVs in BECo's GL 8910 program will be assigned a safety ranking by an expert panel based on risk significance and capability margin. The rankings will be used to develop static diagnostic test frequencies. As a member, the licensee has committed to the JOG to dynamically test two additional valves, the steam supply valve to the high pressure coolant injection system turbine (MO 23012)

and cora spray test bypass valve MO 1400-4A. Dynamic test results from other JOG members will be evaluated and applied at Pilgrim as applicable. The licensee also is installing an automated data acquisition system (SENTRY system) that will provide diagnostic information for the associated MOV each time the valve is operated, Conclusions The licensees' MOV periodic verification program was not yet finalized as of the close of this inspection. However, the GL 8910 program is being closed because further NRC review of this matter will be conducted under GL 96-0 E2.3 (Closed) LER 50 293/97-15: Salt Service Water Pomo inooerability Insoection Scope (37551)

On November 4,1997, BECo discovered that the thermal overload trip settings (amps) of the Salt Service Water (SSW) pump breakers were set too low for degraded voltage conditions arid made a formal NRC notification (EN 33212). This condition was identified during evaluation by engineering. Operators declared the SSW pumps inoperable and entered TS 3.5.8 which required restoration of one loop or be in cold shutdown witnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Inspector follow up of this issue was needed due to the safety significance of the potentialloss of all SSW pumps. The inspector reviewed the documentation and discussed the SSW pump inoperability issue with electrical and mechanical engineers, the SSW system engineer, maintenance, operations, and licensing personnel, Observation and Findinos PNPS has five SSW pumps which provide normal cooling for the reactor building closed cooling water (RBCCW) and for the turbine building closed cooling water (TBCCW) heat exchangers. SSW cooling provides the primary heat sink for the containment cooling subsystem through the RBCCW heat exchangers. SSW pumps "A" and "B" cool the "A" side RB/TBCCW heat exchangers and SSW pumps "D" and "E" cool the "B" side exchangers. The "C" pump is a swing pump that can be valved to supply either sid As followup to the NRC service water self assessment (SWSOPI) followup inspection conducted in the summer of 1997, engineering was confirming the emergency diesel generators (EDGs) loads under loss of offsite power (LOOP) accident conditions. As reported in LER 97-011-00, August 1,1997, BECo identified on July 1,1997, that it is possible for a single SSW pump to be operating with both SSW header division valves (MO 3808 and MO 3813) open. This could leave a single pump operating in a runout condition because of a minimum possible system resistance due to al! four RBCCW and TBCCW heat exchangers in parallel flow paths. The concern was that such a runout

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condition could cause pump cavitation due to inadequate net positive suction head (NPSH)

at the high flow rate. This concern was alleviated, for the short term, by operating with one of the division valves closed at all times in accordance with temporary modification (TM) 97 044, SSW System Lineup Change to Close Either MO 3808 or MO 3813. These issues were addressed in the SWSOPl followup Inspection report 50 293/97 05, issued October 21,199 During the EDG loading followup review, November 1997, electrical engineering became aware of new SSW pump performance curves that showed the brake horsepower (BHP),

especially for the single pump operation supplying one cooling loop, was significantly higher than indicated by the previous performance curves. BECo had shipped the components of a spare SSW pump assembly to their present pump vender (since 1C91),in June 1997, who assembled the pump and performed the flow testing. The new curve, dated August 20,1997, by the vender, showed the BHP for single pump operation at the minimum required SSW flow to both RBCCW heat exchangers (4500 gpm) was 113, significantly above the original pump vendor's value of 96 BHP, and wellinto the 115%

service f actor margin. Electrical engineering realized that under degraded voltage conditions (10% degraded), this increased the fullload motor current to 142 amps. Since the overload heater protection dial was set at 100%, resulting in a minimum trip of about 130 amps and a nominal trip at 153 amps, reliable SSW operation in the one pump at 4500 opm degraded condition could not be assured it was recommended that the overload heater protection dial setting be increased to 115% to increase the minimum trip to about 150 amps and a nominal trip at 176 amp The condition could have caused a SSW pump to trip if one SSW pump had been operating and a degraded voltage condition existed when the 4.16kV on-site power distribution system is being powered by the off site 345kV transmission system. Degraded voltage protection at Pilgrim Station includes alarm and trip functions to alert operators of the condition and protect electrical equipment. The condition is not a concern during a loss of coolant accident (LOCA) with a loss of off site pov.er (LOOP) since the emergency diesel generator would start and power the safety related electrical bustes and hence a degraded voltage condition would not exis With the engineering analysis completed, the licensee declared the five S0W pumps inoperable at 7:40 pm on November 4,1997. In accordance with TS 3/4.5.Blimiting condition for operation (LCO), with both SSW loops inoperable, cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was required, Under Field Revision Notice 97 02 38," Increasing the Settings of the Overload Relay Setting", Maintenance Request (MR) 19702708, * Replace Overload Relays [for SSW Pumps "D" and "E" Breakers)" and Test lin accordance with)

Procedure 8.Q.3 3, *480 VAC Motor Control Center Testing and Maintenance", two spare breakers were modified to set the overload heaters to 115%, bench tested to confirm operability, and installed in the SSW "D" and "E" breaker motor control center (MCC)

cubicals. This work was completed at 3:00 am and the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months shutdown LCO exite The plant was in TS 3/4.5.8 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months shutdown LCO because the second SSW loop was not operable until the "B" and "C" (swing pump) breaker overload heaters were reworke This work was repeated, under MR 197027087,and the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO exited at 8:00 a The "A" SSW pump was out of service for repair of a broken pump shaft. The same breaker modifications were made for SSW pump "A", under MR 197027086,on November 5,199 __, . _ .-_ .

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The inspector reviewed the historical record of past SSW pump breaker thermal overload problems. The following LERs were identifie LER 78 011 SSW pump 208C tripped on thermal overload on 3/17/78. Investigation of the pump breaker revealed that one connector supplying the overload heater relay was loose. The relay was replaced and satisfactorily teste LER 80-037 SSW pump 208B tripped on thormal overload on 7/9/80. No probable cause for the trip could be determined. Tha licensee said this was an isolated even LER 80-049 SSW pump 208B tripped on thermal overload on 8/10 and 11/80 and pump 208E tripped on 8/11 and 17/80. The cause of these trips was SSW pumps being operated at run out conditions. Corrective actions were improved biofouling control and increased overload heater siz Following the SSW breaker thermal overload trips listed above, in 1980 the original H90 overload heater relays were replaced with FH92 models. This raised the minimum trip to about 130 amps and a nominal trip to 153 amps. Since that time, no other breaker overload heater trips were identified and the relay settings were not changed until the November 5,1997 change addressed above. The trip ability of the SSW pumps was routinely tested in BECo's preventive maintenance program, c. Conclusions With a single SSW pump operation per loop and under degraded power conditions, the SSW breaker overload heater protection relays for any pump could have tripped the pump and, therefore, violate TS 3/4.5 B, Operable Core and Containment Cooling Systems. This condition wcs unknown by the licensee prior to receipt and review of the new SSW pump performance testing curve in August 1997. The licensee has determined that this condition was reportable. This condition is considered an apparent violation of design control per 10CFR50, Appendix B, Criterion Ill. (EFI 50-293/9713 07). LER 50 293/9715 is close E8 Miscellaneous Engineering issues (92903)

E (Closed) Violation 50 293/97 02 02:Inadeaunte Deslan Control (Closed) LER 50 293/97 07 Two self disclosing events occurred in May 1997 that rendered safety related systems inoperable due to inadequate design control measures, in reply letters dated June 20, 1997, and July 11,1997, BECo admitted to the violation and committed to several longer term corrective action The first event involved the shutdown of safety related 480/120 voltage regulating transformers during a loss of offsite power event on April 1,1997. The microprocessor control units for these transformers were subsequently modified to disable the under voltage and over voltage shut down functions. Unknown to BECo, these shutdown

, features had been installed by the vendor back in 1991. Shutdown of these voltage regulating transformers placed the plant outside of the design basis as documented in

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Licensee Event Report 50 293/97 07, dated May 1,1997. BECo also initiated actions to provide more rigorous oversight of vendor related activitie The NRC subsequently identified additional concerns with the thoroughness s! errective actions with regard to the control of digital modifications. Further details arc ca..med in NRC Inspection report 50 293/97 12. The NRC enforcement aspects of the additional digital design control concoms were discussed at an enforcement conference held on November 21,1997 in the NRC Region I office. LER 50 293/97 07 and VIO 50 293/97-02 02 are administratively closed since the under and over voltage features were modifind and further evaluation of the control of digital modifications will be avaluated and tracked as part of NRC apparent violation 50 293/9712 01(i.e., eel 50 293/9712 01).

E8.2 (Closed) URI 95 22 01: BECo Corrective Actions to Address Vendor Manual Undates Insoection Scoce (92903)

The inspector reviewed the status of BECo's actions to address the high backlog of open vendor manual change requests (VMCRs). The vendor equipment technicalinformation program status report dated March 2,1995, indicated that there were 195 equipment

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technicalinformation evaluations open with an average age of 441 days. The October 2, 1995, report revealed that there was a total of 176 items open; of these 119 were overdue for a period of between 1.5 to 3.25 years. The failure by BECo to follow NUORG procedure NOP84A4 requirements to ensure all action stems (vendor manual change requests) be completed in a timely manner was treatet as a non cited violation in NRC inspection report 50 293/95 2 Observations and Findinas There are approximately 570 vendor manuals,177 of which are for safety related equipment. Procedure NOP 84A4, " Vendor Equipment Technicalinformation Program,"

requires that all safety related VMCR be entered into the integrated action item data bate (IADB) with a due date of 60 oays from assignment for safety related vendor manuals, or 90 days for non safety related vendor manual Corrective actions taken to resolve the backlog of VMCR included revising procedure NOP 84A4 to simplify the vendor manual change process, in 1996 BECo established a team of engineers to enter / verify entered all vendor manual changes into the lADB and processed the changes to reduce the backlog of vendor manual changes. The ownership of the program was also changed from the Plant Group Manager to the Nuclear Engineering Services Group (NESG) Manage The inspector reviewed the NESG October 199'itrending report and noted that there are 125 open VMCRs. A review of the December 10,1997,IADB revealed that of the open VMCRs,53 were for safety related equipment; several of which have been open since 1993. The due dates for the recent VMCRs were consistent with procedure NOP 84A4 requirements. The inspector discussed the status of the VMCRs with the NESG Manager

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and was informed that BECo intended to place continued focus on the backlog of VMCRs and planned to address the older items by the middle of 1998. The inspector noted that the backlog of VMCR has not contributed to problems with plant equipment. BECo's work control process requires that prior to commencing work on a component all procedures and vendor manuals are required to be verified for current revision at the time work commence Conclusions Corrective actions have been taken by BECo to address the backlog of open vendor manual chang 3s. A downward trend of open VMCRs bas been noted with plans to address the older VMCRs. The open VMCR have not contributed to problems with plant equipmen The failure to process all vendor manual changes in a timely manner was previously determined to be a non cited violation. No additional violations of NRC requirements was noted; therefore, this item is close E8.3 (Closed) URI 96-01 02 Reactor Core Isolation Coolina (RCIC) System inocerability (Closed) LER 50 293/96-Q2 The inspector reviewed the corrective actions taken to address the loss of position indication for RCIC motor operated valve MO 1301 16 during surveillance testing. BECo submitted a voluntary LER to report this even On February 5,1996, the position indication for the inboard steam isolation valve for RCIC was lost as operators slowly jogged the valve. Investigation revealed that the most probable cause was due to the a loose control power fuse. The surveillance procedure included steps to remove and to install the fuse, but the accountability to check for tightness between the fuse holder and fuse was unclea Corrective actions to address loose fuses included developing training for operators, electricians, mechanics, and instrumentation and control technicians regarding the proper handling of fuses / fuse clips, in addition safety-related procedures were reviewed to ensure that when a fuse is pulled, that the fuse is examined, cleaned, and tightened, if necessar The inspector verified that training modules had been developed to discuss this concern, and that a proceduro change was made to the RCIC surveillance ta require sign offs and verification upon re installation of fuses. The corrective actions were determined to 8 s appropriate. The resident inspector conducted an in office review of this event repoa and concluded that this event was not required to be reported under 10 CFR 50.73. No violation of NRC requirements existed. This item is close E8.4 (Closed) LER 50 293/95 08; Documented that valve MO 120180, a primary containment system / reactor water cleanup isolation valve, was improperly wired. The condition would not allow the valve to automatically travel beyond the 96 to 98 percent closed position. Upon identification, the applicab!c technical specification action was entered and the condition immediately corrected. Investigation revealed that the root cause was a personnel error during a modification installation performed in ',993. The other isolation valvo in the line remained operable and the subject valve was able to be

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closed via a manual control switch. BECo concluded that the piping penetration remained operable. The broader issue regarding wiring discrepancies was tracked by ur. resolved item (URI) 50 293/95 15 01. The URI was subsequently closed in NRC Inspection Report 50 293/07 04, This licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vil.B.1 of the NRC Enforcement Policy. (NCV 50-293/97 13 08). This LER is close E8,5 (Closed) LER BO 293/9612: documented that the reactor had been operated slightly above 1998 megawatts thermal (MWT)in excess of the time allowed by technical specifications. An omission existed in the calculation for determining core thermal power (CTP). The maximum CTP that could have occurred would not have exceeded 199 MWT The condition was identified by the licens e when investigating a er.dition riescribed in the operating experience computer network. BECo indicated .t this condition is bounded by the loss of coolant analysis wh2ch assumes an ir,itial reactor power level of 102% (i.e.,2038 MWT). This event was clossd based on an in office review by the resident inspector. This licensee identified and corrected vlotation is being treated as a Non Cited Violation, consistbra with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50 293/97 13 09). This LER is close E8.0 { Closed) Unresolved item 50 293/93 22-01: Pressure lockina and thermal bindina of nate valves. The licensee had not completed its evaluation of safety-related gate valves for susceptibility to pressure locking and thermal binding. The inspectors reviewed calculation M-600, "MOV Pressure Locking and Thermal Binding Evaluation," Revision 3, dated February 15,1996. The calculation contained the results of the licensee's susceptibility evaluations under normal operating, emergency, shutdown, and testing conditions. DECO identified twelve MOVs that were susceptible to pressure locking and modified them to preclude the condition either by installing valve bonnet vents or relief valves, or drilling a hole in the valva disk. Three valves susceptible to thermal binding were modified to reduce inertial closing (wedging) thrust and to increase mctor actuator capability. The licensee documented these actions in a letter dated February 23,1996, responding to GL 05 07," Pressure Locking and Thermal Binding of Safety Related Power-Operated Gate Vt.!ves." The licensee's evaluations and modifications were sufficiently timely and complete to warrant GL 8910 program closure. Final review of this issue at Pilgrim, including the need for any further licensee activities, will be documented in a safety evaluation report (SER) on GL 95 07 issued by the NRC Office of Nuclear Reactor Regulatio E8.7 (Closed) Unresolved item 50 293/93 22 02: Ambient temocrature effects on AC fnotors. When this item was opened in 1993, BECo had not completed its evaluation of motor actuator capability in response to Limitorque Technical Update 93-03, which provided information concerning elevated ambient temperature affects on motor output torque. Several MOVs preliminarily were evaluated as having marginal capability when these effects were taken into account, and valve modifications were being considere The inspectors reviewed calculations PS-132 (133,134, and 135), " Electrical Performance and Stroke Timing Evaluation for Priority 2 (3,4, and 5) AC MOVs," and Nuclear Engineering Department Work Instruction (NEDWI) 428, " Generic Letter 8910 MOV l

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31 ,

Calculation Methodology For The Electrical Evaluation of AC Powered MOVs." The calculations made certain conservative or bounding assumptions including: (1) initial cable temperatures were assumed to be 90 degrees Centigrade, (2) peak post accident temperatures inside the primary containment were assumed regardless of what period during the transient the MOVs were called upon to operate, (3) a 10 degrce Centigrade :

motor temperature rise was assumed regardless of actual valve stroke time, (4) the elevated temperature motor current correction factors specified in Limitorque Update 93 03 were not used, resulting in lower (i.e. more conservative) motor terminal voltages, and (5)

minimum avellable motor torque was determined based on the ratio of minimum motor termint.1 voltage to rated voltage raised to a power of 2.2. The licensee's methodology .

was consistent with Limitorque Corporation's recommendations and current industry standards (e.g. Institute of Electrical and Electronics Engineers (IEEE) Standard 1200 1996,

"lEEE Guide for Motor Operated Valve (MOV) Motor Application, Protection, Control, and Testing in Nuclear Power Generating Stations."

Regarding the valve specifically discussed in this unresolved item, Loop B core spray test bypass valve MO 1400-4B,the licensee enhanced the output capability of the valve by upgrading the motor, and Installing a stem with increased lead and a new gear se Fourteen additional valve motors were upgraded to a higher torque rating or operating speed in part as a result of high ambient temperature effect As documented in NRC Inspection Report 50 293/97 05,the licensee recently completed a new post accident containment temperature profile that increased peak amblent temperature from 331 to 334 degrees Fahrenheit. The licensee stated that the electrical calculations were being revised to reflect the new peak temperature. The ir.spector concluded that the 3 degree rise would not reduce MOV capabilities significantly. No violations of NRC requirements were identified. This item is closed based on the licensee's implementation of limitorque update 93 0 E8.8 flgyjgyr of Uodated Final Safety Analysis Reoort A recent discovery of a licensee operating its facility in a manner contrary to the updated final safety analysis report (UFSAR) description highlighted the need for a review that compares plant practices, procedures, and/or parameters to the UFSAR descriptions. While performing the inspections documented in this report, the inspector reviewed motor-operated valves in the recirculation, core spray, residual heat removal, and high pressure coolant injection systems and corresponding sections of the Pilgrim UFSAR. The inspectors verified that the UFSAR wording was consistent with the observed plant practices and procedure E8.9 Criticality Accident Reauirements NRC inspection Report 50 293/9711 closed unresolved item 50 293/97 02 04. This issue involved the failure to have in pace either a criticality monitoring system for storage and handling of new (non-irradiated) fuel or an NRC approved exemption to this requirement contained in 10 CFR 70.24. A final review of the issue was performed during this inspection period.

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The NRC has concluded that a violation of 10 CFR 70.24 existed. The NRC has also determined that numerous oths! licensees have similar circumstances that were caused by confusion regarding the continuation of an exemption to 10 CFR 70.24 originally issued prior to lasuance of the Part 50 license. After considering all the factors that resulted in these violations, the NRC has concluded that while the violation did exist, it is appropriate to exerciso enforcement discretion for Violations involving Special Circumstances in accordance with Section Vil B.6 of the " General Statement af Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG 1600. Pending the emendment to 10 CFR 70.24, further enforcement action will not be taken for f ailure to meet 10 CFR 70.24 provided an exemption to this regulation is obtained before the next receipt of fresh fuel or before the next planned movement of fresh fue LE PLANT SUPPORT R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 (Closed) Violation 97 03-02: Radiological Controlled Area (RCA) Boundarv Door An RCA boundary door in the boiler room was found tied open without proper controls as required by procedure 1.3.114," Conduct of Radiological Operations." Immediate actions included securing the door and checking other RCA doors for a similar problem. No other problems were identified. in the reply letter to the violation, dated August 21,1997, BECo admitted to the violation. Severallonger term corrective actions focused on improvements to the training program for waste control technicians. During this inspection period, the inspector independently checked several RCA boundary doors and identified no concerns. The inspector considers violation 97-03-021s close R4 Staff Knowledge and Performance in RP&C

R4.1 Radioloalcal Work Practice Insnection Scone (71750)

The inspector monitored maintenance workers' radiological work practices during the replacement of main steam safety relief valve RV 203 3 Observation and Findinas Workers adhered to the radiological work permit and had the proper anti-contamination clothing and dosimetry. During the performance of the work activity an interference was encountered and additionai rigging and tools were required. Maintenance workers demonstrated good ALARA by exiting the drywell while these items were obtaine During the performance of the work activity a mainteiiance supervisor, who was observing the work activity, answered a cellular phone while in the orywell without removing his rubber gloves and placed the phone against his face. After the conversation was completed, the phone was placed back inside of the anti-contamination clothing. No personnel contamination was monliored upon exiting the radiological controlled crea. The mspcctor informed the maintenance supervisor and radiation protection personnel about

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this unacceptable radiological work practice. A problem report was generated and the supervisor counseled. This failure c. institutes a violation of minor significance and is being treated as a non cited violation, consistent with Section IV of the Enforcement Polic (NCV 971311) Conclusiont Overall, proper radiological controls were demonstrated by the maintenance craf t during -

replacement of the main steam safety relief valve. One poor practice was identified when a maintenance supervisor demonstrated poor contamination control work practices which had the potential to cause a personnel contamination by retrieving his cellular phone from inside of the anti contamination clothing without removing his rubber gloves and placing the receiver against his f ac V. MANAGEMENT MEETINGS X1 Exit Meeting Summary BECo representatives were informed of the scope and purpose of this inspection at an entrance meeting on November 10,1997. Observations and findings were discussed with licensee representatives during the course of the inspection and at an exit meeting conducted on December 9,1997. No proprietary materials were reviewed during this l Inspectio X3 Management Meeting Summary The NRC conducted an enforcement conference on November 11,1997, with BECo management in the NRC Region 1 office. Attached to this report as Attachment 111 is a t handout provided by BECo. The issues involved NRC Inspection Report 97 05 and 971 X4 Review of UFSAR Commitments A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for additional verification that licensees were complying with Updated Final Safety Analysis Report (UFSAR) commitments. For an Indeterminate time period, all reactor inspections will provide additional attention to UFSAR commitments and their incorporation into plant pracoces and procedures. While performing inspections discussed in this report, inspectors reviewed the applicable portions of the UFSAR No inconsistencies were note I

.

INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls in identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observation -

IP 62707: Maintenance Observation IP 71707: - Plant Operatint IP 71750:- Plant Support Activities IP 82301: Evaluation of Exercises for Power Reactors

IP 927_00: Onsite Followup of Written Reports of Nontoutine Events at Power Reactor l Facilities l lP 92901: Followup Operations IP 92902: Followup Maintenance l IP 92903: Followup Engineering IP 92904: Followup Plant Support IP 93702: Prompt Onsite Response to Events at Operating Power Reactors

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ITEMS OPENED, CLOSED, AND UPDATED Pilgrim Nuclear Power Station NRC Inspection Report 00 293/9713 Ongned 97 13 01 VIO Failure to follow work control procedures - (nataisd Mong teiay !n ATWS cabinet 97 13 02- VIO Rector Vessel Flange Metal Temperature Indication l 97 13 03 IFl Valve factor assumptions

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l - 97 13 04 IFl Recirculation pump isolation valves

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l 97 13 05 lFl Stem friction coefficient l-97 13 07 eel Breaker overload heater protection relays for single SSW pump operation Closed 93 22 01 URI Pressure locking and thermal bindin9 of g6te valves 93 22 02 URI Ambient temperature effects on AC motors 95 22 01 URI BECo Corrective Actions to Address Vendor Manual-Updates

- 96 01 02 URI Reactor Core isolation Cooling (RCIC) System inoperability 96 80 02 URI Work control issues 97 01 01 VIO Procedure adherence to feed water system off normal procedure and locked valve list 97 01 02- VIO Inadequate corrective actions for three operator work arounds and also for a missed standby switchyard battery surveillance 97 02 02 VIO Inadequate design control -

97 03 02 VIO Radiological controlled area (RCA) boundary _ door 95 08 LER Primary containment sys isolation valve unable to close -

95 12 LER Core thermal power exceeded limit

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96 02 LER Loss of position indication of RCIC MOV during surveillance test Closed items, et LER Procedure adherence to feed water system off normal procedure and locked valve list 61 07 LER Losses of off site power during storm / shutdown 07 15 LER Salt service water pump inoperability 97 25 LER Shutdown due to inoperable MSIV 97 26 LER Automatic reactor scram 97 13 06 NCV Design calculations 97 13 08 NCV Improper wiring of RWCU valve 97 13 09 NCV Exceed core thermal power limit 97 13 10 NCV Missed maintenance PWT 97 13 11 NCV Minor HP error work practice 97 13 12 NCV Minor violation - outdated thread sealant issued to the job-site

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UST OF ACRONYMS USED ALARA As Low As Is Reasonably Achievable

- APRMs . Average Power Range Monitors BECo Boston Edison Company CFR Code of Federal Regulations CRD Control Rod Drive

! CS Core Spray

[ EP Emergency Preparedness l EPIC Emergency and Plant Information Computer i

ESF Engineered Safety Feature gpm gallons per rninute I&C Instrumentation and Controls IFl Inspection Follow Up ltem IR Inspection Report LER Licensee Event Report MG Motor Generator MR Maintenance Request NCV Non Cited Violation NOV Notice of Violation NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NWE Nuclear Watch Engineer PNPS Pilgrim Nuclear Power Station PR Problem Report RHR Residual Heat Removal RP Radiological Protection SALP Systematic Assessment of Licensee Performance SRO Senior Reacto;' Operator T Temporary Modification TS Technical Specification UFSAR Updated Final Safety Analysis Report WWM Work Week Manager I

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Pilgrim Nuclear Power Station NRC Predecisional Enforcement Conference November 21,1997 G

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PILGRIM STATION VISION At Pilgrim Station, our passionate commitment to excellence and each other make us successfulin all we d We have a safe work environment that values and encourages teamwo*. We provide our customers with safe, reliable, competitively priced, and environmentally clean electricit .

11/20/1997,1100

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.I KEY MESSAGES e We understand the FSAR needs to be upgraded and we are committed to do thi i e We understand Pilgrim needs to dearly document its design 1 basis information and this effort is underway, e We will ensure plant procedures contain accurate and complete design informatio * We understand the need for conservative reporting for design basis issues and the need to lower our threshold for 50.59 evaluation * We will continue to improve our corredive action process and its oversight functions to ensure timely identification and effective resolution of issues, e We will educate our oversight groups to provide critical assessment of our performance against these expectation , .

PRE-DECISIONAL ENFORCEMENT CONFERENCE AGENDA Design Basis and FSAR H. V. Oheim Process or Program Issues Design Control J. P. Gerety 50.59 J. P. Gerety 50.72/50.73 N. L. Desmond FSAR Update (50.71) N. L. Desmond EQ (50.49) J. P. Gerety Oversight and Corrective Action H. V. Ohelm Regulating Transformers J. P. Gerety Enforcement Perspective N. L. Desmond Closing Summary L. J. Olivier Closing Executive Remarks T. J. May 11/20/1997,1100 3-4

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DESIGN BASIS AND FSAR H. V. Oheim

.f DESIGN BASIS INFORMATION

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f RECOGNIZED AS A CRITICAL NEED e Design Basis information Program undenuay e Objectives . . .

-Organize, verify, validate Pilgrim design basis-Reconstitute appropriate information-Ensure plant procedures contain complete, accurate design inforrration-Create a living document supported by a sound document maintenance program 11/20/1997.-1100

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DESIGN BASIS INFORMAhlON RECOGNIZED AS A CRITICAL NEED (cont.?

e Program oased on . . .

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-NRC documents

- NEl Guidelines-Industry best-practice _ programs i

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PROGRAM APPROACH AND SCOPE e Cross-Functional Team Approach-Experienced Pilgrim engineers, ,

operators, maintenance personnel-Experienced contractor support e 40 to 60 systems will be included,10 to 15 topical reports-1997 - 2 systems,1 topical report-1998 to 2001 - 16 to 18 systems per year,4 to 6 topical reports 11/d0/1997,1100

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PROGRESS ON 1997 SCOPE

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Program framewc.t. procedures complete oWork on 2 safety systems underway e System / topical report priority set by maintenance rule systems list and IPE '

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DESIGN BASIS PROGRAM END RESULTS IMPROVE STATION OPERATION e User friendly documented design basis . . .

-Provides valuable tool to operators, engineers, and maintenance personnel-Expands understanding of plant design

~ Provides a solid foundation for plant changes e End products will be living documents 11/20/1997,1100 9-10 l

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DESIGN BASIS PROGRAM END RESULTS

IMPROVE STATION OPERATION i; cont.ll e Criticallinkages between design, licensing basis, and plant procedures I

e information will be used to clarify and ,

improve FSAR FSAR PLAYS KEY ROLE IN SWOPI, FUTURE OPERATIONS

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e Several SWOPl issues tied to technical differences e Origin of differences linked to FSAR

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FSAR PLAYS KEY ROLE IN SWOPI, FUTURE OPERATIONS (cont.)

e Reinforces Pilgrim's assessment of

FSAR condition

. -Assessment began June 1996-Purpose was to ensure plant conformed i to FSAR l -Results indicated in-depth review

needed-Process described in 50.54f response

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SCAQ PROBLEM REPORT !SSUED

TO ANALYZE FSAR ISSUES e PR issued in 2/97

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o Root cause analysis completed 9/97 e issues found with . . .

-programs supporting FSAR maintenance

-personnel knowledge-expectations and oversight

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e Results form basis for FSAR improvements

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11/20/1997,1100 13-14

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FSAR IMPROVEMENT CONCURRENT

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WITH DESIGN BASIS PROGRAM I

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be verified and clearly documented

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e Design basis program includes FSAR j information validation i

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e Management of FSAR will be further proceduralized and integrated into

. interfacing processes e Update' timeliness has been proceduralized e independent oversight will monitor effectiveness ofimprovements

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SCOPE AND SCHEDULE BEING

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FINALIZED

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e Full scope and schedule will be docketed in 50.54fletter update by

12/19/97 e Progress will be reported in LTP

., f DESIGN CONTROL J. P. Gerety 11/20/1997.1100 17-18

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'. DESIGN CONTROL ISSUE

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e NRC cited four examples of-inconsistencies between plant analysis and operating procedures-Operation with SSWinlet temperature to RBCCW above 65'F-No procedure step to isolate non-essential RBCCWloads following a LOCA-instrument uncertainty not incorporated into RHR flowrate indication-Individual load estimates for the EDG in procedure did not match design calculation values DESIGN CONTROL . . . CAUSES e Lack of a clearly documented design basis e inadequate technical review of procedures for instrument uncertainty e inadequate tracking of end use of engineering calculations e Translation of analytical values from

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. design calculations to operating procedures has not always been effective 11/20/1997,1100 19-20 I

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DESIGN CONTROL . . . CORRECTIVE

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ACTIONS COMPLETED eThe PNPS licensing and design basis have been revised to resolve the USQs

[7/3/97]

e EDG procedure revised [7/3/97]

e RHR procedure change evaluated using

, 10CFR50.59 [11/6/97]

e SSW pump tested for cavitation impact

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e Run RHR hydraulic model with Hx valve

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i DESIGN CONTROL . . . CORRECTIVE l ACTIONS REQUIRED

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o Perform RBCCW calculation taking no

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credit for SSW for the first 10 minutes

[11/21/97]

e Revise RBCCW hydraulic calculation to

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loads not isolated [11/28/97]

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DESIGN CONTROL'. . . CORRECTIVE l ACTIONS REQUIREDicont.)

e Formalize the process to translate

analytical limits into usable values forl

, use in operating procedures as part of

, the design basis program [2/27/98]

.e Review normal, abnormal and- .

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emergency procedures for safety related systems _to idendfy those which involve-

. operator actions based on indication to verify consistency with FSAR and design

bas, [1/15/98]

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ACTIONS REQUIREDfcont.)-

!' e Train on performing a technical review of a procedure. The training will be

. directed at making clear the expectation

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that a technical . review requires that l values given in operating procedures be

{ tracked back to a basis document and

- include required uncertainties. [1/31/98]

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.. DESIGN CONTROL . . . CORRECTIVE-ACTIONS TO PRECLUDE RECURRENCE e Complete detailed design' basis program for risk.significant systems at PNP The design basis shall also include the

licensing basis and a technical review of corresponding system operating procedures.

i DESIGN CONTROL . . . CORRECTIVE ACTIONS TO PRECLUDE RECURRENCE (cont.)

e Interim Correction Actions:

'A design basis program has been initiated. The areas selected for-immediate review based on the PNPS IPE risk significance are:

- DC System,

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- Residual Heat Removal System, .

- A design basis report for FSAR .

Chapter 14 Accident Analysis 11/20/1997,1100 25-26 I

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. DESIGN CONTROL . . . CORRECTIVE ACTIONS TO PRECLUDE RECURRENCE

i; cont.)

e improve the process that captures and maintains engineering calculations to ensure procedures are reviewed whenever a calculation is revised

[12/31/97]

Q 50.59 J. P. Gerety 11/20/1997,1100 27-28 i

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110CFR50.59. . . SAFETY EVALUATIONS eThe NRC noted four examples where-safety reviews or safety evaluations did not reach appropriate conclusions:

-Operation above 65' SSW temperature-Crediting containment pressure for NPSH-isolation of non-essential RBCCW loads-RHR flow procedure change 10CFR50.59 . . . CAUSES (

e The licensing basis for PNPS for the use of containment overpressure and allowable salt service water temperature was ambiguous, leading to different interpretations 11/20/1997,1100 29-30 t

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. 110CFR50.59. . . . CAUSES (cont.)

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' e Individuals'did not always understand what constituted a change' to the facility

- as described in the FSAR e Historically, the values given in the FSAR -

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for system performance have been a mixture oflimits, minimum values,

. nominal values and analytical value :There has not been any standard applied to values liste _ , #,,4-10CFR50.59. . . CORRECTlVE ACTIONS

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COMPLETED o Amendment No.173. allows use of

- containment overpressure and a Salt Service Water temperature of 75'F e Safety Evaluation 3118 was approved by TORC on 11/6/97, to allow the revision of RHR procedure 11/20/1997,1100 31 32 f

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COMPLETED (cont.) .

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e Individuals who perform, review, and audit 10CFR50.59 evaluations were.

l instructed on what constitutes a change

- to the facility or a change to the i . procedures as described in the FSAR and i

the purpose of a Preliminary Evaluation Checklist (PEC). [11/20/97]

10CFR50.59. . . CORRECTIVE ACTIONS REQUIRED

e Document the PNPS licensing basis e Revise PNPS procedures to provide

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additional guidance to determine when a 10CFR50.59 evaluation is required

[1/31/98]

e Schedule third party assessment of 50.59 process [2nd Qtr 1998]

e Training will be provided annually through

ESP program

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. #g 50.72/50.73-REPORTABILITY N.L.Desmond

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,ff REPORTABILITY ISSUE e NRC cited six examples of conditions which they determined were reportabl One of.those was reported, but later than expected.

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REPORTABILITY . . . CAUSES i

PNPS personnel. inappropriately used operability to deterrnine reportability-Operability' evaluations based on ,

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engineering judgment versus completed I analysis (five examples).

-The FSAR does not clearly distinguish design basis documentation from design information (one example)

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.REPORTABILITY . . . CAUSES (cont.)

e Human Error- failure to consider historical temperatures for EDG temperature issue e Human Error- misinterpretation of EO guidance to maximize drywell cooling

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REPORTABILITY . . ; CORRECTIVE-

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ACTIONS COMPLETED

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e Estabiished additional guidance toi ensure appropriate' repo ting of outside design basis and unanalyzed conditions e Applied th.e new guidance to NRC issues and made required 50.72 repod ;[11/4/97)

' e' Included new reporting guidance in the PR procedure and the regulatory affairs

' work instruction [11/18/97)

REPORTABILITY , . . CORRECTIVE ACTIONS COMPLETED (cont.)

e Instructed regulatory, operations, and engineering supervisors and managers

on new reporting guidance and the importance of conservative reporting

[11/20/97)

e Updated Regulatory Affairs work

. instruction to ensure timely reportability evaluations [11/18/97).

11TdO/1997,1100 39-40

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REPORTABILITY . .'CbRREbTIVE

. ACTIONS REQUIRED

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e Review of open prs forwhich we are operating under an engineering evaluation using new reporting guidance

[ scheduled 12/21/97]

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REPORTABILITY . . . CORRECTIVE ACTIONS REQUIRED (cont.?

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e Upgrade the FSAR to clearly distinguish design basis documentation from design

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N. L. Desmond l

S.E.1638 NOT CAPTURED IN.FSAR

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UPDATE e issue: Modified recirculation system f

pipe insulation not reflected in the FSAR (1985)

e Probable Cause:

-SE concluded no FSAR impact-SE missing ORC signature (FSAR update process incorporates only approved SEs)

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i 50.71. . . POSSIBLE CONTRIBUTING

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CAUSES e Containment analysis results presented

, in confusing manner-e Use of SE revision numbers

< e Maintain existing FSAR level of detail e Failure to question relevance of SE1638 during FSAR update

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50.71. . . CORRECTIVE ACTION e Corrective Action Completed-SE1638 was superseded by SE2971

[1996)

-missing information incorporated into 1996 FSAR update (rev.19)

-SE revisions no longer permitted e Corrective Action Planned-establish FSAR quality standards-upgrade FSAR: licensing / design bases 11/20'1997,1100 45-46 u -

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EQ (50.49)

J. P. Gerety

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10CFR50.49: ENVIRONMENTAL QUALIFICATION e From 1988 until early 1996, Pilgrim was outside the design basis drywell

- temperature profiles and could have exceeded established Environmental Qualification design temperature limits

for equipment [ identified 1/26/96]

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e Based on information provided by GE for the 1987" model, the analyst

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results.

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10CFR50.49. . . ACTIONS COMPLETED

l e'All equipment covered by 10CFR50.49 has been qualified [3/16/96)

i e Current process is to. require all l computer code inputs and assumptions .

be reviewed by BECo

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CORRECTIVE ACTION ISSUES H. V. Oheim

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CORRECTIVE ACTION ISSUES

. Corrective action process did not resolve some significant problems identified in the SWOPl self assessment in a timely, effective manne /20/1997,1100 51-52

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CAUSES e Once operability was assuredLsome issues did not receive appropriate emphasis

.e Appropriate priority and management attention was not always applied CORRECTIVE ACTION . . .

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incomplete actions were closed e The "special" status of SWOPl items-carried into follow-up actions,

-ci rcumventd e corrective action program

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CORRECTIVE ACTIONS COMPLETED e Transfor of Corrective Action Team to QA to allow us to focus on closure

l e Open SWOPl action items converted to problem reports, increased attention and trending CORRECTIVE ACTIONS TO BE COMPLETED e Close remaining 4 SWOPI prs (original scope was 114 items)

e Complete design and licensing basis projects

Strengthen management attention on backlog open issues and their completion 11G0/1997,1100 55-56

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gm .,u.._______, _.-- - - 2c1 . . - - m -~ sr x. m. m. a

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REGULATING TRANSFORMERS J. P. Gerety -

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[4-BACKGROUND . . . VOLTAGE REGULATING TRANSFORMERS eThe 120/480 vac regulating transformers tripped during LOOP on

.4/1/97-Event was repoded to NRC

---Design control and vendor oversight issues captured in reply to NRC violation 97-02-02 11/20/1997,1100 57-58 i

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REGdLATINGjRANSFORMER ISSUES >+ .. - '

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e Regulating Transformers (120/480Lvac)? .

1 Linstalled in'1992.withou* recognizing the;

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. . Texistence'of the autom'aticinternal; o

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.undervolta'g e trip; .

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c tel Regulating Transformers m'odified in

+ " 41997 without evalisating' digital >

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Ico'mpoIYent upgrade requirements .

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REGULATING TRANSFORMERS . . .

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OPERATIONAL-lMPACT e Both regulating transformers are close to _

_the control room and easily accessible (Power can be restored within minutes)

. * Regulating circuit can be bypassed to -

allow continued use of the transforme ,

o Abnormal procedures direct bypassing of regulating circuit, if required, and were tsuccessfully implemented during the April 1, event 11Tdo/1997,1100 59-60

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REGULATING TRANSFORMERS . . .-

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.CAUSES-

. e The purchase specification did not require the transformers to continue operation during all voltage transients

[1992]

e Transformers were purchased through a

_

third party dedicator who did not notify BECo that an automatic undervoltage trip existed [1992]

e Modification to transformers was not recognized as a digital upgrade [1997]

f l REGULATING TRANSFORMERS . . .

f CORRECTIVE ACTIONS COMPLETED e Design change to remove the automatic undervoltage trip implemented .[4/12/97]

e Operability evaluation completed for operation while software verification is completed [10/9/97]

No digital modifications allowed until procedural guidance issued 11TdO/1997,1100 61-62

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REGULATING TRANSFORMERS . . .

CORRECTIVE ACTIONS REQUIRED 4

o Complete verification of software

[12/5/97]

e EMI testing :[12/10/97]

e10CFR50.59 issued [12/18/97]

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l REGULATING TRANSFORMERS . . .

ACTIONS TO PRECLUDE RECURRENCE e Develop a procedure for performance of-digitalmodifications [1/15/98)

e Procedure on preparing procurement specifications was revised to require suppliers to notify BECo of all automatic

.funrf,ns installed in new equipment

[8/1/9/]

11/20/1997,1100 6344 I

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! - ENFORCEMENT PERSPECTIVE

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N. L. Desmond

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? _ .4 MANY ISSUES SELF IDENTIFIED DURING SWOPI SELF ASSESSMENT

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o RHR procedure vs. analysis e Isolation of non-essential P3CCW loads e 65 degree SSW temperature issue

- e EDG ambient temperature issue

- e Other issues identified during BECo resolution-Error in containment' analysis-Use of containment pressure for NPSH

11/20/1997,1100 65-66

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SYSTEMS. CAPABLE OF. PERFORMING

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SAFETY FUNCTIONS..

l eThe following issues did not impact :

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capability to: perform safety function

-RHR. procedure vs. analysis ,

- 5 d gree SS tm rat re-SSW single failure vulnerabilit Error in containment analysis

.Use of containment pressure for NPS

' -EDG - third party review confirms

- cperability - analysis in progress

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e Regulating transformer issue resulted in minimalimpact on operating crews

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ENFORCEMENT PERSPECTIVE

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I e No conflicts with technical specifications e.No actual or potential impact on public e Need to upgrade FSAR selfidentified

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l4 11/20/1997,1100 67-68 f

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. IMMEDIATE CORRECTIVE ACTIONS TAKEN e EDG procedure corrected e Safety evaluation prepared for RHR procedure changes-e 50.72 reports completed and LERs in progress e Reporting guidance modified

IMMEDIATE CORRECTIVE ACTIONS TAKEN (cont.)

e Personnel instructed on importance of

.conservative sa ft e y rev iews and reportability reviews e SSW pump testing conducted 11i20/1997,1100 69-70 t

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..fg CORRECTIVE ACTIONS ARE COMPREHENSIVE o Broad Scope programs initiated-FSAR upgrade-Design basis program

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-Licensing basis documentation e 50.59 process fundamentally sound

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h CLOSING SUMMARY

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L. J. Olivier Vice President- Nuclear Operations 11/20/1997,1100 71 72

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KEY MESSAGES

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o Pilgrim team committed to sofo conservative operations and full compliance with the regulatio * We understand the FSAR needs to be upgraded and we are committed to do thi * We understand Pilgrim needs to deariy document its design basis informat'on and this effort is underwa * We will ensure plant procedures contain occurate and complete design information, e We understand the need for conservative reporting for design basis issues and the need to lower our threshold for 50.59 evaluation ;

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KEY MESSAGES (cont.)

We will continue to improve our corrective action process and its oversight functions to ensure timely identification and effective resolution of issue * We will educate our oversight groups to provide critical assessment of our performance against these new expectations.

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We will update 50.54(f) letter by December 19,1997,

We will update NRC on our progress during the first ;

quarter of 1998, 11/20/199/,1100 73 74

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EXECUTIVE CLOSING REMARKS T. J. May Chairman and CEO

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11/20/1997,1100 75-76 9

v t e Inspection Report - Pilgrim - 1997013
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 80(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000293/1997013

Text

REGION I

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