IR 05000293/1997011

Insp Rept 50-293/97-11 on 970914-1110.Violations Noted:Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20197H248
Person / Time
Site:	Pilgrim
Issue date:	12/11/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20197H238	List: IR 05000293/1997011 ML20197H234 ML20197H241
References
50-293-97-11, NUDOCS 9712310229
Download: ML20197H248 (27)

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- U.S. NUCLEAR REGULATORY COMMISSION

REGION I

License No.: - DPR 35

Report No.: - 97 11_  ;

. Docket 'No.: 50-293 Licensee: - Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 ,

Facility: Pilgrim Nuclear Power Station i Inspection Period: September 14,1997 November 10,1997

- Inspectors: _ R. Laura, Senior Resident inspector

< R. Arrighi, Resident inspector T. Fish, DRS Operations Examiner A._ Wang, NRR Project Manager ,

Approved by: C. Cowgill, Chief '

Reactor Projects Branch No. 5 Division of Reactor Projects

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.a 9712310229 971211 i PDR- ADOCK 05000293 9 PDR

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EXRCUTIVE SUMMARY-Pilgrim Nuclear Power Station NRC Inspection Report 50 293/97-11 Ooerations -

• Generally, the conduct of operations was professional and safety consciou Operators closely moniiored an increase tailpipe temperature from safoty relief valve 3D which indicated pilot valve seat leakage. (Section 01.1)

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• The recirculation pump trip event was handled very well by the operating crew by following procedures to place the plant in a stable and safe condition. The root cause and corrective actions were still under evaluation at the completion of the inspectun period. (Section 01.2)

• Investigation into the cause of the 125 VDC ground experienced during "D' EDG starts was not aggressively pursued, and the troubleshooting plan delayed due to the batteries being cross tied. (Section 02.1)

'* Three individual licensed operator errors, involving the tagout preparer, tag hanger and independent verifier, resulted in tagging an electrical breaker in the wrong position. Collectively, these errors resulted in a violation of the PNPS tagging

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program. (Section 04.1)

• A licensed operator requalification program inspection was conducted October 14 17,1997. The training staff developed acceptable written and operating tests, and administered and evaluated the tests effectively. A notable -

strength of the program was the training evaluators, who were particularly good at determining how well operators and crews mastered training objectives. Their high standards for performance were reflected in their critical and objective assessment Based on these favorable attributes, th6 inspectors concluded that the requalificanon program was very good and contributed to safe plant operatio (Section 05.1) .

• BECo promptly notified the NRC of an intemal safety concern involving an operator log discrepancy in 1989. A good initial BECo response was evident by hiring the

- services of an independent investigative service and limiting the licensed duties of '

any operator who may have been in the control room at the time. Minimum shift manning was not compromised. (Section 08.1)

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Maintenance

• A planned limiting condition for operation (LCO) maintenance outage on the "A" emergency diesel generator did not go as smoothly as past LCO maintenance issues. The NRC identified two problems including an inadequate tagout for the K-li i

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103A starting air compressor and slow corrective action for a broken coolant expansion tank Hi/Lo level alarm dating back to 1993. The inadequate tagout was considered a second example of a violation of the PNPS tagging process. (Section M1 1)

• The failure of the "A" salt service water (SSW) motor shaf ts was due to f atigue f ailure induced by improper pump assembly. Based on the satisfactory alignment checks of the "B", "C", "D", and "E" SSW pumps and the metallurgical evaluation performed on the f ailed shafts, the failure mechanism was not a generic concern for the other SSW pumps. NRC review of the corrective actions taken after the first shaft failure remains unresolved. (M2.1)

Enaineerina

• A missed technical specification (TS) surveillance was averted due to self identification by BECo and issuance of an emergency TS amendment. This error resulted from miscommunication within the engineering department and inadequate oversight of the Inservice test program. (Section E1.1)

• BECo was susceptible to the suppression pool bypass path under certain plant evolutions (torus inerting). Prompt actions were taken by the ODM upon identification of the issue to prevent placing the plant in a condition outside of its design. (Section E1.2)

Plant Sucoort

• Two instances of poor radiological practices were observed including an overflowed protective clothing receptacle and spillage of water from a dewatering rig outside of a posted contamination area. Minor cleanliness issues were obseived in the Torus room. (Section R1.1)

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TABLE OF CONTENTS EXECUTIVE S UMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . li _ .

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Summary of Plant St atus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -

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.l . O PE R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 - Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Recirculation System Loop "B" Pump Trip . . . . . . . . . . . . . . . . . . 3 ,

l 02K Operational Status of Facilities and Equipment ....................4 02.1 Emergency Diesel Generator (EDG) . . . . . . . . . . . . . . . . . . . . . . 4 '

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04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 5 04.1: (Open) VIO 50 293/97-11-01: Tagging Errors . . . . . . . . . . . . . . . 5 05 Operator Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 05.1 Requalification training (LORT) program . . . . . . . . . . . . . . . . . . . 7 ,

-08- . Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 .

08.1 (Open) IFl 50 293/9711-02: Internal Employee Concern ' . . . . . . 8'

08.2 - (Closed) IFl 50-2 9 3 /96 80-01 . . . . . . . . . . . . . . . . . . . . . . . . . . 9 11. M AI NT E N A N C E ' . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 10 M1 Conduct of Maintenanca . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 M1.1 (Open) VIO 50 293/97-11-01: General Maintenance ......... 10 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . 12 '

M.2.1 (Open) URI 50-293/97-11-03: Salt Service Water (SSW) Pump Motor Shaf t Shear; (Closed) ....,............,............12 lil . EN GI N EERING . . . ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E1- Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E Inservice Testing (IST) of Reactor Building Closed Cooling Water (RBCCW) Check Valve 30-CK-432 . . . . . . . . . . . . . . . . . . . . . 13 E.1.2 Potential Suppression Pool Bypass Path . . . . . . . . . . . . . . . .. 15 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 E.8.1. (Closed) LER 5 0-2 9 3 /97-0 9 . . . . . . . . . . . . . . . . . . . . . . . . . .16 E.8.2 (Closed) LER 50-2 9 3/9 5-1 1 -01 . . . . . . . . . . . . . . . . . . . . . . . . 16

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E.8.3. (Closed) LER 50 2 9 3/9 5-06 . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 E.8.4 (Closed) URI 50-2 9 3/9 7 0 2-04 . . . . . . . . . . . . . . . . . . . . . . . . 16 E.8.5 (Closed) URI 50-293/94 26-01: Diesel Generator Turbo Assist Solenoid Valve Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 -

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I V. PLA NT S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 7 R1 _ Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 17 R1.1 Tour of radiological controlled areas . . . . . . . . . . . . . . . . . . . . . 17 V. M AN AG EMENT MEETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 -

X1_ _ Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 18 X4- Review of UFSAR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 iv

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.: lNSPECTION PROCEDURES USED . . . -. . . . . =. . . . . . . . . . . . . . . . . . . . ., . . . . . . . . 20 -.

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LITEMS OPENED, CLOSEDi AND UPDATED ~: . . . . .:. . . . . . . . . . . . '. . . . . . . . . . . . -. 21I

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L LIST OF ' ACRONYMS USED . . . , . . . . . . . . . . . . . . . . . . . . . . . .'~ . . . . . . . . . . . . . . 2 2 - j :

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REPOR1 DETAlld Summarv of PlartStatus Pilgrim Nuclear Power Station (PNPS) began the period at approximately 100% reactor '

power. Significant [.ower reductions of interest to the NRC are described belo On September 19, operators lowered reactor power to approximately 53% due to the loss of the *A* first polqt feed water heater. A f alse heater hl watee level alarm and isolation occurred due to a steam leakage from a level transmitter gasket. Af ter corrective maintenance was performed on September 20, operators returned the unit to full powe On November 5, an inadvertent trip of the "B" reactor recirculation pump, during restoration from a surveillance test, caused power tv decroase to approximately 70%.

Su'sequently, operators further lowered reactor power to approximately 35% to restart the

"D" reactor rocirculation pump. The unit was returned to full power on November 6. A formal NRC notifiention (ENS 33220) was made pursuant to 10 CFR 50.72(b)(2)(li)to report the inadvertent ESF actuation. Further details of this event are discussed in Section 01.2 of this repor The plant operated at approximately 100% reactor power during the remainder of this inspection perio IJfTBAIl0RS 01 Conduct of Operations'

01.1 fieneral Comments 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 any observed alarms with the operators and verified that they were aware of any lit alarms and the reasons for them. For example, on September 28,1997, operators closely monitored an increasing tailpipe temperature from safety rCief valve (SRV) 3D which indicated pilot valve seat leakage. The requirements of techrtical specifications (TS) 3.6.D.3 and 3.6.D.4 for SRV seat leakage were well understood and followed. A positivo synergy was observed between the SRV system engineer and control room operators. An operability evaluation was promptly completed ensuring continued operability for SRV 3D as well as compensatory measures such as increasing the set point of the downstream tailpipe temperature alarm. The inspector regularly observed the plant manager and operating department manager tour the control room to closely monitor any degraded equipment condition 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 topics. .

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The inspector reviewed the latest INPO training accreditation report dated September 1997 which contained no significant safety issues. The report indicated that overall BECo training performance was positiv The inspector accompanied nuclear plant operators (NPOs) during daily rounds in the turbine building and reactor building. The tours are performed in accordance with procedure 2.1.16," Nuclear Plant Operator Tour." The purpose of the tours are to check the plant on a routine frequency for abnormal conditions during all modes of plant operation. The inspector noted that the NPOs had the proper tour sheets with the latest revision available and in use during the toar. The tF)s were knowledgeable of equipment status and monitoring equipment performance as required. During the tour of the reactor building, the inspector noted a light fixture tucked in an electrical cable raceway. This had gone unnoticed by the NPOs on their rounds. Upon identification of the concern, the light fixture was immediately remove Anomalies noted during tours were discussed with the nuclear watch engineer (NWE). The inspector noted two important degraded equipment conditions (i.e., worn recirculation motor generator (MG) exciter brushes and reactor vessel water level reference leg keep fill sptom relief valve seat leakage PSV 263 A/B)in the plant which had work request tags (WRT) attached indicating the problems were identified and entered into the work control syste With regard to the leakage of at least 100 drops / minute downstream of valve PSV 263 A/B, the inspector questioned the operability of the keep fill system since only extremely sinell flow rates (approx. 006 to .01 gpm) of water get injected into the reference le The operations staff initiated problem report (PR) 97.9692 to address the inspector's question. Operatora subsequently took actions to resent the leaking relief valve which .

stopped the seat leakage prior to exceeding the 14 day tracking LCO. In reviewing this matter, the inspector determined that senior reactor operators did not contact the system engineer in a timely manner and were generally non consarvative by not declaring the keep fill system inoperable and entering the 14 day tracking LC As for the worn brushes on the exciter of the recirculation system pump MG set, the inspector questioned the acceptable length of continued MG set operation without replacing the brushes. One possible consequence of the worn brushes could be the initiation of a recirculation pump trip ar.d plant transient. The brushes were worn to a red band wear marking as identified during a monthly electrical preventive maintenance chec The system engineer later informed the inspector that the wear was considered normal wear and test and that satisf actory operailon of the brush was assured up to 3/8 inch into the red marking. As a conservative measure, the rsplacement of the brushes was moved up from the initial scheduling during the planned April 1998 shutdown outage to a planned down power on November 21,1997. The operations department manager informed the inspector that an enhanced review of WRTs will be performed to ensure that problem reports were initiated when require Other specific events and noteworthy observations are detailed in the following section .

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01.2 Recirculation System Loon "B" Puri;o TJja  : Inspection Scoce (71707)

On November 5,1997, at 1810 hours0.0209 days <br />0.503 hours <br />0.00299 weeks <br />6.88705e-4 months <br />, an unplanned closing of tha recirculation system loop *B" pump discharge valve MO 202 5B occurred during the performance of a  !

surveillance of the emergency diesel generator "B" initiation circultri. The closing of the valve resulted in the automatic trip of the "B" recirculation pump, consequent decrease in

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reactor core flow and reactor power, from 100 percent to approxiraately 70 percent, and reactor operation in the caution zone of the reactor core power flow relationship. The resident inspector responded is the site to independently monitor operator performance in restoring the plant to a stable condition, and followed up on the licenset's investigation into the cause of the event, Observetions and Find!Dat Licenced operator response included the entry into procedure 2.4.17," Recirculation Pump Trip," and the insertion of control rods to achieve reactor operation at less than the 100 percent load line. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limitine condition for operation fLCO) was entered in accordance with the f acility operating licence condition 3.E. Recirculation loop "A" pump speed was reduced and reactor power was lowered to 35 perecnt power in accordance with procedure 2.2.84, " Reactor Recirculation System." The *B" recirculation pump was restarted and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO terminated at 2225 hours0.0258 days <br />0.618 hours <br />0.00368 weeks <br />8.466125e-4 months <br />. Correct procedure adherence was noted. The inspector noted that the Plant Group Manager also responded to the event and was in the control room monitoring operator performanc The *B" recirculation pump automatically tripped during restoration from the performance.,

of surveillance 8.M.2 2.10.8.2," Diesel Generator "B" Initiation by RHR." This test is required to be performed once per operating cycle. The "B" emergency diesel generator was saGafactority started and shut down, an operator reset the low pressure coolant injection (LPCI) loop select logic, and the residual heat removallogic switch was taken to reset when valve MO 202 5B went close The Nuclear Watch Engineer directed that the surveillance test be secured and contacted the lastrumentation and Control supervisor to investigate possible causes, investigation revealed that closure of the valve was due to the failure to obtain proper reset of the LPCI loop select logic (two relays remained energized). The relays had been energized as part of the surveillance test and should have de energized when the operator pushed the "LPCl Loop Select" reset push button, When the button was pushed again the relays de-energized as expecte Review of LPCI loop select circuit logic and relay testing revealed no problems with the circuitry or push button. Preliminary investigation by BECo determined that the cause of the event was a procedure weakness. The procedure did not include a step to verify that all relays were de energized prior to resetting the logic circuitry at the completion of the test. There is no indication in the cor. trol room that the LPCl logic is reset, it was not the intent of procedure 8.M.2 2,10.8.2 to test the proper function of the two LPCIloop wiect

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logic relays. Corrective actions willinclude revising applicable procedures to incorporate lessons learned,

' ConclusioD The recirculation pump trip event was handled very well by the operating crew by following procedures to place the plant in a stable and safe condition. The root cause and corrective actions were still under evaluation at the completion of the inspection perio Operational Status of Facilities and Equipment 02.1 Emproency Diesel Generator (EDG) Insoection Scop.e (71707)

On October 23,1997, a 125 VDC ground occurred during the performance of procedure PNPS 8.9.1, " Emergency Diesel Generator and Associated Emergency Bus Surveillance."

This was the third instance of a ground being experienced durin0 the : tart of the "B" ED Grounds also occurred on July 21,1997, and again on October 16,1997. The inspector teviewed BECo actions taken relative to this issue, _ Observations and Findinag Problem reports were generated for each of the grounds experienced on the 125 VDC system. A review of each event revealed that the grounds were spurious, occurring during the initial start of the "B" EDG, for a duration of two to three seconds. Operators considered the "B" EDG operable based on the alarm immediately clearing and the equipment operating as designed. The engineering evaluations stated that a single ground '

condition (momentary / solid) will not prevent the EDG from performing its intended design functio The inrpector reviewed problem reports dating back tu January 1997 and did not identify any ott er instances where a ground occurred during the start of an EDG, A review of the

"B" EDG performance data revealed that there wer: nyn se::cessful EDG surveillance runs without experiencing a ground between the July 21 and October 16,1997, EDG run No corrective actions were taken by DECO af ter the ground experirnced on July 21. This was based on the subsequent satisfactory performance of the EDG monthly surveillance r Jn; and that the DC system is const'mtly monitored for grounds and alarms in the control

- room te ensure that appropriate corrective actions are initiated. BECo directed that a direct cause be initiated after the second occurance of the ground on October 16,1997,in addition, maintenance request (MR) # 19702545 was written to instrument the DC system in an attempt to locate'the ground. The direct cause e. valuation and MR had not been completed when the third ground occurred seven days late Maintenance request # 19702545was available to be implemented on November 21, 1997, after the troubleshooting plan was generated, it was scheduled to be performed

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during the next scheduled *B" EDG monthly surveillance run on November 24,199 However, due to the "A" and *B" batteries being cross tied, the MR was deferred. The MR was re scheduled to be performed during the next monthly EDG run ir. December. The inspector questioned DECO regarding the delay in troubleshooting the ground due to the uncertainty of continued EDG performance. BECo subsequently scheduled the work to be performed during the current plant shut down (November 22),in part, due to the insr.ectors concer far,Lghtg]gQ investigation into the cause of the 125 VDC ground experienced during "B" EDG starts was not aggressively pursued, and the troubleshooting plan delayed due to the batteries being cross tie Operator Knowledge and Performance r'4.1 [Ooon) VIO 60 293/97-1101:Lanina Errors Insocction Scone f 71707)

The inspector reviewed the circumstances that led to the electrical breaker for the 125 VDC backup battery charger breaker (i.e., B1021) being tagged and independently checked by two licensed reactor operators in the cpon position when in f act the breaker was closed. This problem was identified by an off going senior reactor operator (SRO) who was passing through the cable sprooding room. The breaker was immediately opened as required by the tagout. The inspector interviewed the tag second checker, assistant operations department manager ar,d reviewed the apparent cour,e analysis (ACA) for .

related problem report (PR) 97.290 Observations and Findinag Tagout T97 46162 was prepared with the " normal" position of electrical breaker 1021 listed as closed, which was the first error, since the actual normal position for B1021 is open. The ACA stated that the third SRO who prepared the tegout did not verify the normal position for 81021. The inspector noted that the ACA focused heavily on the errors made by the tag hanger and independent verifier, but did not explore the reason for the tagout preparer error for listing the incorrect " normal" position. The inspector it,arned that a computer database was utilized to develop the tagout instructions. The computer database incorrectly listed B1021 as normally close vice open. The tagout preparer did not detect this error made by the computer database. Although not discussed in the ACA, the inspector confirmed that the tagout database error had been corrected. The tagout preparer violated step 6.2, General Instructiont., of procedure 1.4.5. revision 43, "PNPS Tagging Procedure," which specifies that tageuts shall be prepared using controlled documents such as piping and instrumentatio1 drawings (P&lDs) and electrical prints. The inspector expressed concern about other potential operations computer tag database deficiencies. The licensee acknowledged the inspector's concern and stated that the computer data base would be reviewe .

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Alto, the inspector observed that the nuclear watch engineer (NWE) authorized the

" approval to isolate" section for togout T97 46162 and also did r,ot detect the error in the normal position for B102 Based on reviaw of the tegout sheet which listed breaker B1021 as normally close, the tag hanger concluded that the breaker had to be manbulated to the open position. The physical orientation of 81021 is approximately six feet abovo floor level with a horizontal motion to change the breaker position. The tag hanger inadvertently closed B1021 and affixed the danger tag. T07 46162, tag no. 3, specified that B1021 be tagged in the open position. This action by the tag hanger violated procedure 1.4.5, section 6.7.2[5),

Tagout Placement, which specifies that a qualified individual will reposition devices as directed by the tegout assignment sheet. The ACA adequately discussed the error made by the tag hange Lastly, the second reactor operator performing the independent verification failed to identify that 81021 was tagged in the wrong position. This action violated procedurn 1.4.5, step 3.0112), independent Verification, which specifies that the verifier shall independently locate the subject component and confirm that the component has been lef t in the stato intended by the instruction. When questioned by the inspector, the independent verifier indicated that he did not effectively use the prescribed self checking process. The ACA adequately documented the details of the actions of the verifier including a check of security log records to confirm that the tag hanger and verifier entered the proper plant areas as expected during the implementation of the tagout instruction No potential falsification issues we..e identified during this revie The inspector noted the positive aspect of the problem identification made by the off poing SRO and problem report initiation by the on watch NWE. Also, the problern was self identified and corrected prior to ti e start of any maintenance activities. However, the inspector remained concerned since three separate licensed operator made errors during the preparation and installation of the tag. Based on the multiple errors and a weakness in the apparent cause analysis, enforcement discretion was not applied to this licensee identified violation. These errors represent the first example of a violation (50 293/9711-01) of procedure 1.4.5, "PNPS Tagging Procedure". Section M1.1 of this report discusses a second example of a tagging violation. The apparent cause analyeis focused heavily on the tag hanger and verifier but did not evaluate and document in depth the error made by the tagout preparer, Conclusion Three individual licensed operator errors, involving the tagout preparer, tag hanger and independent verifier, resulted in tagging an electrical breaker in the wrong position. While licensee identified, collectively, these errors resulted in a violation of the PNPS tagging progra __ _

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05 Operator Training and Qualification .

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05.1 Reaustification teainina (LORT) orocram Insoection Scope (71001)

I During the week of October 13, inspectors evaluated the Pilgrim licensed operator requalification training (LORT) progra Observations and Findinas Reaualification examinations: The inspector reviewed simulator scenarlos and noted tha

. training staff incorporated system and procedure changes, and industry events into simulator examinations, when applicable. For example, the staff developed and  !

- administered a new scenario which tested a recent revision to the Alternate Reactor i Pressure Vessel (hPV) Depressurization procedure. TFe revision directed operators to open all four safety relief valves (SRVs) to rapidly depressurize the RPV, a change from the previous requirement to open at least three SRVs for rapid depressurization. Also, based

' on their review of tha dynamic scenarios, the inspector determined that the operating tests adequately sampled various performance abilities for reactor and senior reactor operator The inspector also reviewed the 1996 written examination and the associated sample pla They noted the sample plan selected toples concerning system changes, procedure changes, and current events and that the written examination tested items associated with these topics. The examination also adequately sampled the operators' knowledge, skills,- '

and abilities needed to perform reactor and senior reactor operator dutie Examination adrninistration oractices: The inspector observed the training staff evaluate one operating crew (six operators) on the simulator. The staff administered two scenario The evaluators, which included a senior representative from the operations department, were very good and graded to high standards for acceptable operator performance. They detected individual and crew errors and responded appropriately. For example, during one :

scenario involving secondary containment control, the evaluation team noted that the nuclear operations superintendent (NOS) misread a step in the emergency operating procedures (EOPs) and prematurely depressurized the RPV when plant conditions did not require such action. The evaluators also observed a reactor operator cause a large level transient due to his poor control of thn feed water system. The evaluators f ailed both operators, and removed thern from the watch bill pending successful completion of a remediation program. The training staff also correctly noted that, overall, the crew

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exhibited less than-optimal compliance with off-normal procedures while concurrently executing EOPs. Accordingly, the staff targeted this competency as an area for increased future training and evaluatio The inspector also observed the training staff administer job perforrnance measures (JPMs). As noted during the scenarios, the training evaluators demonstrated critical observation skills and appropriately graded errors several operators committe ,

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Finally, throughout the inspection, the inspector did not detect any indications of exam colnpromis Feedback system: The inspector reviewed a representative sample of licensed operator comments and noted the training staff effectively implemented feedback where appropriate. For example, the inspector noted a comment from an operator challenged the correctness of an exam answer. The feedback system tracked this comrmat and the staff changed the answer ke Remedial trainina oroaram: The inspector evaluated the remediation provided to operators to correct deficient performance and noted the training had corrected weaknesses. For example, one operator demonstrated poor performance in the area of procedure knowledge and familiarity. Accordingly, the training staff tal6ored a remediation program that focused on this area. Subsequent testing of the individual showed that his proficiency had improve Ooerator license conditions: The inspectors reviewed the records for the re activatlen of 1 operator licenses for 1995,1996 (no re activations), and 1997 and noted the licensed opera' >rs had met watch standing proficiency requirements. Also, based on a review of six operator medical records (roughly 10% of all operator licenses), the inspector determined that physical exams were being performed and documented properly, Conclusions The exam for operater requalification training had good quality, including good sampling of various areas. BECo ensured licensed operators satisfied the conditions of their licenses. .

The training staff effectively revised the requalification program based on the operators'

performance in the plant and on requalification examinations. Finally, the training evaluators were particularly good at determining how well operators and crnws mastered training objectives. High standards for performance were reflected in their critical and objective assessments. Based on these f avorable attributes, the inspector concluded that the requalification program was very good and contributed to safe plant operatio OS Miscellaneous Operations issues 08.1 (Open) IFl 50 293/971102: Internal Emolovee Concern Insoection Scooe (71707)

On September 22,1997, the inspector was informed by BECo management of an internal nuclear safety' concern involving a potential falsification issue of a control room log in 1989. The specific concern involved the inadvertent actuation of a safety system while the reactor was shutdown. The inspector reviewed the immediate actions initiated by BECo senior management to evaluate the concern including potential impact on plant operatio , -- , - .- , -,

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- BECo issued a site wide news release, issued a press release and made a formal NRC notification pursuant to 10 CFR 50.72 (b)(2)(vi) for offaite notifications. NRC Region I issued Preliminary Notification 197 061 and responded to several press inquiries. A formal BEco investigation was initiated by hiring an independent firm to review this matte Current licensed plant operators who may have been present in the control room on the day of the event in 1989 have been temporarily reassigned to non licensed duties pending '

the outcome of the investigation. The inspector observed no significant adverse effect on shif t manning and independently verified compliance with technical specification Table 6.2- i 1, Minimum Operating Shift Crew Composition. The inspector was informed that at the l

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conclusion of the BECo investigation a report will be issued to document the finding Pending BECo completion of the investigation and NRC review of the findings, these issues will remain as an Inspector follow up Item 50 293/971102 j e Conclusions l

BECo promptly notified NRC of an internal safety concern involving an operator log discrepancy in 1989. A good initial BECo response was evident by hiring the services of an independent investigative service and to limit the licensed duties of any operator who may have been in the control room at the time. Minimum shif t munning was not compromise .2 (Closed) IFl 50 293/96 80 01:The potential was identified that certain recirculation

, pump trip scenarios may result in reactor power and flow combinations that place the plant in an unauthorized area of the power to flow map. This item was opened pending NRC

- evaluation of BECo's corrective actions to this condition. BECo staff determined that at low flow conditions inaccuracles are introduced into the core flow indication. Also, the indication circuitry introduces non linearities and additional errors. Therefore, it is very difficult to establish any calibration criteria at low power / low flow conditions. Instead, it  !

was more desirable to have accuracy in the high power /hlgh flow condition since the plant typically operates under these conditions. Although this event had not occurred at Pilgrim, BECo staff nevertheless revised operating procedures 2.1.14, " Station Power Changes,'

2.4.1/ Recirculation Pumpfs) Trip," 2.2.84, * Reactor Recirculation System," and revised the power to flow map to show operators that the stability boundaries extend below natural circulation. The inspector noted the plant staff completed the revisions and determined the changes were appropriate. This item is closed.

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i 11. MAINTENANCE I M1 Conduct of Maintenance M 1.1 (Onen) VIO 50 293/971101: General Maintenance Insnection Scoon (61726,62707)  ;

i Using inspection procedures 61726 and 62707, the inspector observed portions of selected maintenance and surveillance activities to verify proper calibration of test -

Instrumentation, use of approved procedures, performance of the work by qualified personnel, conformance to limiting conditions for operation, and correct system reMoration 1 follo*ving maintenance and/or testing. Portions of the following activities were observed:

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• P9701041 "A" EDG quarterly preventive maintenance (PM) i

• P9700252 "A" EDG preventive maintenance ';

• P9600808 Change oilin K 103A compressor

• P9600807 Change oilin K 107A compressor i e19702237 Replace EDG jacket water e.9702435 Repack "B" salt service water pump < Observations and Findinas A planned limiting condition for operation (LCO) maintenance outage was performed for the

"A" emergency diesel generator (EDG) which entered the unit into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> sht.tdown pursuant to technical specification (TS) 3.9.B.3 and 3.5 F.1. The inspector observed tha Work generally progressed well Due to the presence of moisture in the compressor oi sump, the maintenance mechanics and supervisor expanded the work scope for the PM on the starting air compressor K 103A. Mechanics first tried flushing the moisture from the oil sump and then removed an access cove to wipe the moisture using rags. The removal of moisture in the sump was noted to be a positive action. Also, an emergent item identified just prior to the outage was added to the "A" EDG outage scope to replace a '

leaking air receiver drain valve which had to be cut out and a new valve welded in plac in general, the inspector observed good performance by both the mechanics and '

supervisor. Two problems were identified by the inspector during the "A" EDG LCO outage.-

First, the inspector noted that the local "A" EDG alarm C 103B A1, Coolant Expansion

- Tank Hl/Lo was broken as evidenced by the pulled alarm assembly. A review of the maintenance history revealed that this alarm failed its functional test in May 1993. This deficiency was entered into the work control system as evidenced by the existence of -

MR19301726. Work planning personnel informed the inspector that the high level float

switch was broken and a part procurement issue existed. The float manufacturer has since

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gone out of busines Initial engineering efforts per Field Revision Notice (FRN) 96-04-14, dated February 7, 1996, did not work due to inadequate clearances. The work control department manager

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Indicated that this repair was initially included in the planning for the "A" EDG LCO maintenance window this assessment period, but was removed due to the required engineering work not being complete. Subsequently, engineering issued FRN 90-04115 on October 9,1997 during this inspection period and parts werc ordered. Orsce the parts are obtained, an EDG outage will be required to install the new parts since the coolant expanslor, tank must be draine The inspector determined that DECO's efforts to repair the coolant expansion tank Hl/Lo level alarm have been slow. The inspector noted that this deficiency was longstanding, and expressed concern to licensee management about operators being required to take compensatory actions for an extended period of time while repairs are complete A second problem was oSserved on October 7,1997, which involved inadequate mechanicalisolation for P9700252. On the previous shift (night shift) mechanics disconnected a small copper pressure sensing line connected to K 103A, starting air compressor. The inspector identified that air was bleeding out from the disconnected copper sensing line which in effect was bleeding down the "C" starting air receiver, T-146C. The inspector subsequently identified an untabehd manualisolation valve in the smallinstrument sencing line from K-103A was not shut and tagged as part of the boundary tagout. The inspector discussed this observation with shif t operators and tag n was added to tegout T97 6144 to close the open boundary valve in the sensing lin On October 10,1997, the maintenance supervisor initiated problem report 97.2997 to evaluate the cause of the tagging error. At the end of this inspection period on November 11,1997, operations department management had not yet completed the apparent cause analysis for the tagging erro The inspector reviewed tagout T97 6144,which was initially prepared, in part, for preventive maintenance request P9700252 to perform PM on the "A" EDG P9700252 references section 1 of procedure 3.M.3-61.5, EDG PM. Step (3)(c) in Section lA includes work instructions for the K 103A unloader. In ne past, tagouts for the EDGs bleed down the air start receivers. Tagout T97 61-44 did not bleed down the air i, tert receivers. The inspector identified the second example of a violation (50 293/971101)of procedure 1.4.5, PNPS Tagging Proceduie, section 4.2.1 which specifies that valves must be set in the proper position to ( olate the equipmen The LCO outage for the "A" EDG lasted 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> vice the planned 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The work control manager conducted a lessons learned review to identify opportunities for improvement. The lessons learned review determined that better coordination of plant groups by the work wsek manager was required. With few exceptions, the inspactor considered the review was done in detail and demonstrated the ability to be self critical, Conclusions A planned LCO outage on the "A" EDG did not go as smoothly as past LCO maintenance issues. The NRC identified two problems including an inadequate tagout for the K 103A starting air compressor and also slow corrective action for a broken coolant expansion tank

- Hl/Lo level alarm dating back to 199 .

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12 l M2 Maintenance and Material Condition of Facilities and Equipment M.2.1 (Open) URI 50 293/971103: Salt Service Water (SSW) Pumo Motor Shaf t Shear:

(Closed) IFl 50 293/97-08-01 Insoection Scone (62703)

On October 19,1997, the motor drive shaf t for the "A" SSW pump sheared while in service. This was the second f ailure of the "A" SSW motor shaf t in a two month period; ths shaft had previously failed on August 26,1997. The inspe:: tor reviewed the cause of these shaf t f ailures to determine the potential generic applicability to the other four SSW pumps, and the corrective actions implemented after it * .haft f ailure, Observations and Findinas The second (replacement) "A" SSW pump motor shaft f ailure occurred at the shaf t thread next to the couplirg between the motor shaft and the pump shaft. The first shaft had failed at the keyway and not the threaded area. Both f ailures were similar in that: they occurred while the pump was in service, the fracture surf ace appearance was similar, and there was no apparent damage to the gib key or upper keyway, in addition, both shaf ts were trom the same lot number; with the first one being in service for 18 months and the second for over four years. The shafts were sent to Mass Materials for destructive and non-destructive examination to determine the cause of the f ailure Laboratory testing results of the first "A" SSW motor shaft confirmed that the material and chemical properties were normal. The metallurgical evaluation revealed that the f ailure was due to a mixed mode f atigue/ tensile mechanism. The fatigue failure was the result of cyclic loading that occurs due to pump vibration as well as pump starts. The c'yclic loading caused the initial defect / crack to propagate. Preliminary results of the second shaft indicate that there may have been a pre-existing crack in the shaft threads and that the failure was fatigue due to bending. Material properties of the second shaft also appeared to be norrna Af ter the second "A" SSW pump shaft f ailure, BECo examined the "A" pump and identified nignificant misaljnment between the motor and pump. A gap of approximately 0.08 inch between the pump motor and motor support stand was identified. This resulted in misalignment of the pump motor and pump which caused excessive shaft deflection and subsequent fatigue failure. Alignment checks of the other four SSW pumps were performed. This included an inspection for gaps at the motor stand and measurements of the pump shaf t angularity, offset, and runout. Results of these checks were satisf actor BECo investigation of the motor and motor stand misalignment revealed that there was a unique configuration of the "A" SSW pump motor stand. Additional weld pads had been installed to take up clearance due to metalloss (corrosion) from old oump motors. The

"A" SSW pump motor was replaced with a new motor in 1994; however, when the pump was replaced, the normal motor dimensions were restored and the weld pads did not allow proper fit up. The existence of the weld pads was indicated on the motor stand drawing;

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the drawing apparently was not used during the motor replacement. Based on this Information, the problems experienced with the "A" SSW pump motor shafts were unique-to that pump and IFl 50 293/97 08 011s closed. The inspector notes that more comprehensive follow up actions were taken af ter the second shaft failure such as

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removing and inspecting the pum BECo initiated a root cause evaluation which had not yet been completed at the end of this inspection period. The adequacy of the corrective actions taken af ter the first "A" SSW pump motor shaf t breakage will be assessed af ter BECo completes the root cause evaluation. This will remain as an unresolved item (50 293/97-1103).- Conclusions The f ailure of the "A" SSW motor shafts was due to fatigue failure due to improper pump ,

assembly. Based on the satisfactory alignment checks of the "B", "C", "D", and "E" SSW pumps and the metallurgical evaluation performed on the f ailed shafts, the f ailure mechanism was not a generic concern for the other SSW pumps. NRC review of the corrective actions taken after the first shaft f ailure will remain unresolved, t

E ENGINEERING E1 Conduct of Engineering E Inservice Testina (IST) of Reactor Buildina Closed Coolina Water (RBCCW) Check Valve 30 CK-432 Inspection Scone (37551)

On September 18,1997, BECo identified a refueling outage reverse flow exercise (surveillance test) for check valve 30 CK 432 that would be overdue by more than 25% on November 2,1997. The surveillance test should have been performed during the cycle 11 refueling outage. This check valve is required to be tested in accordance with BECo's inservice valve test program every refueling interval. The inspector reviewed the conditions leading to the failure to perform the required IS Observations and Findinas inservice code testing activities are required to be performed in accordance with Technical Specification (TS) section 3.13, " Inservice Code Testing," and 10CFR50.55a(f)," Inservice Testing Requirements." The test frequency for the reverse flow exercise is every two years; with a technical specification allowed extension of up to 25 percent. The surveillance test of valve 30 CK 432 was due to be performed on May 2,1997, and its grace period expire November 2,1997. The check valve is also required to be seat leak testen (LLRT)in accordance with 10CFR50 Appendix J. The licensee had been grantid perm,ssion tc. credit the Appendix J testing (two year frequency) to satisfy the ruve,w . 0W testlig requirements. BECo therefore incorporated both the Appendix J testing and the-reverse flow exercise into one test procedure and tracked them by the same master surveillance tracking program (MSTP) node. -

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On October 4,1996, TS Amendment No.167 was approved by the NRC to allow BECo to implement 10CFR50 Appendix J, Option B. This option allowed BECo to extend the LLRT frequency for valve 30-CK 432 from two years to 60 months based on satisf actory valve leakage performance. As a result of this test frequency change, the licensee changed the node for this surveillance in the MSTP to reflect the now test frequenc The IST coordinator was aware of the change to the LLRT test frequency and the need to develop separate procedures to perform the IST reverse flow check for eight check valves during outages when the LLRT is not performed. Due to miscommunication and inadequate follow through witHn the engineering department, valve 30 CK 432 was inadvertontly omitted from the newly developed procedures. As a result, the reverse flow check for this velve did not get performed during the cycle 11 refueling outage, which ended April 1997. A subsequent review L/ BECo of the IST records revealed that this was the only valve affected by the license amendment that was not tested in the refueling outag Valve 30 CK 432 is an outboard primary containment isolation valve for a system required to be in survice during plant operation, it provides isolation to non safety related drywell equipment cooled by the RBCCW system. Thu performance of the check valve closure test for valve 30 CK 432 during plant operation requires the isolation of both the "A" and "B" loop drywell coolers and the recirculation pump seal water and motor lube oil cooler The licensee determined that performing this test on line would present an unacceptable plant risk (potential plant transient, accelerated equiprnent wear, or a reactor trip on high drywell pressure) and requested by letter dated October 10,1997, a notice of enforcement discretion (NOED) to delay testing of this valve until the next plant shutdown. BECo subsequently submitted an emergency TS amendment request to allow deferral of the surveillance; with the test being performed no later than the 1998 maintanance outag The inspector reviewed the LLRT test results for valve 30-CK-432 and the valve maintenance history to determine overall valve performance. The LLRT test resultc were well within allowed leakage criteria, and there has been no recent maintenance history that indicates unacceptable valve performance. The inspector also verified that the LLRT and reverse flow exercise are being tracked by separate nodes in the MSTP. The emergency TS change was approved by the NRC on October 30,1997. The next scheduled maintenance outage is tentatively scheduled for April 1998, Conclusions A missed TS surveillance was averted due to self identification by BECo and issuance of an emergency TS amendment. This error resulted from miscommunication within the engineering department and weak oversight of the IST pr:, gra ._

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E.1.2 Potential Suporession Pool Bvoass Path Scope (37551)

General Electric 10CFR Part 21 notification No. SC97 4 dated October 15,1997, provided information concerning the potential for suppression pool bypass paths created by a single active failure that could create conditions that were outside of the analyzed containment response. The inspector reviewed BECo's design to determine if the Pilgrim plant was susceptible to this concern, Observations and Findinas Another nuclear utility identified that the control cabling for both the drywell and suppression chamber at their plant was contained in the sume cable raceway. Due to a postulated single f ailure (of multiple hot shorts in the same raceways both of the inboard containment isc!ation valves could be energized to open, allowing the drywell and wetwell atmospheres to communicate. Following a postulate Loss of Coolant Accident (LOCA),if this condition were to occur, it would result in bypass leakage of the postulated LOCA environment around the suppression pool, thus losing the beneficial effects of both the quenching of the blowdown energy and the scrubbing of any fission products. BEco wrote a problem report to review this issue for applicabilit The inspectors' review of BECo plant drawings revealed a potential bypass path between the drywell to the suppression chamber through the containment atmosphere control system; specifically, the nitrogen makeup and purge system. Further review and discussion with BECo staff revealed that the plant is susceptible to suppression pool bypass only during torus inerting. The plant is placed in this lineup approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to taking the mode switch to the run position during a plant startu In a normal valve lineup at 100 percent power, the nitrogen purge path is isolated by two air-operated valves (AO-5035A and AO 5036) and a manually-operated varve (9 HO 117).

If a f ailure occurred that opened the two air operated valves, the path would remain isolated by the closure of the manual operated valve. A hot short that would energize valves AO 5035A and AO 5036A was not credible because the cables running to the valves are in physically separated conduit. However, if relay RPWA1 f ailed to energize, or if its contact 12 failed to open on a primary containment isolation signal, both air operated valves would remain open. Therefore, BECo was susceptible to suppression pool bypass when valve 9 HO 117 is open. Valve 9 HO 117 is only open while inerting the toru Review of the nitrogen make up path revealed that in addition to two air operated valves, there are two check valves in series which would isolate the torus and the drywell. The inspector verified that the subject check valves were part of and tested for leak tightness as part of BECo's Appendix J program. Therefore, this lineup would be adequately isolated in the event of a single failure that resulted in the two air-operated valves failing ope Upon identification of the potential suppression pool bypass path, the Operations

' Department Manager (ODM) generated a tracking licenseo condition of operation to prevent

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openir,g of valve 9 HO 117 until the condition is resolved. BECo made procedural changes to resolve this concern and dotermined that this condition was reportable to the NRC pursuant to 10 CFR 50.7 : Conclusion BECo was susceptible to the suppression pool bypass path under certain plant evolutions .

(torus inerting). Proriipt actions were taken by the ODM upon identification of the issue to prevent placing the plant in a condition outside of its desig ES Miscellaneous Engineering issues l E.8.1 JClosed) LER 50 293/97-09:The reactor core isolation cooling (RCIC) system tripped on over speed during a surveillance test, and on two subt.equent occasion. juring c the troubleshooting evolution. Investigation revealed that the cause of the initial over s speed was the result of a gear change made to the RCIC system flow test return valve  ;

during the refueling outage. The applicable RCIC procedures were not identified as impacted by the gear change. The other two trips were the result of wiring changes performed during the outage that made the test valve a seal-in circuit in the closed direction. Tripping of the RCIC system on over speed was previously discussed in NRC Inspection Report 293/97 02 and a violation of design control was issued. This LER is clo e E.B.2 { Closed) LER 50 293/9511-01: This LER documented a loss of power to the RCIC flow control circuitry due to an intermittent RCIC inverter failure. The inverter was .

replaced and the system restored to normal. Testing by BEco and the vendor were unable to identify eny defects or anomalies with the inverter. Discussions with the system ,

engineer revealed that no further trips have been experiences as a result of inverter failures. This LER is close E.8.3 (Closed) LER 50 293/95 06: This LER documented the RCIC system was inoperable due to a failuie in the RCIC turbine steam supply valve investigation revealed that either a loose wire or a failed auxiliary relay in the seal-in relay within the breaker which supplies - ,

power to the valve motor operator. The wires . vere checked for tightness, the auxiliary relay replaced, and the valve was satisfactorily tested. Testing of the auxiliary relay revealed no discrepancies. Discussions with the system engineer revealed that no further trips have been experienced as a result of steam supply valve. This LER is close E.8.4 [Cjosed) URI 50 293/97 02-04: By letter dated October 10,1597, the NRC issued Information Notice (IN) 97 77," Exemptions from the Requirements of Section 70.24 of Title 10 of the Code of Federal Regulations." In this notice the NRC stated that "the staff does not intend to take further enforcemeist action for failure to meet 10 CFR 70.24 provided that licensees obtain an exemption from this regulation before the next receipt of-finsh fuel or before the next planned movement of fresh fuel." In addition, the staff

.~ intends to withdraw the previously issued violations BECO has stated that an exemption

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will be requested before the end of 1998. Therefore, based on lN 97 77 and the planned ,

exemption request by BECo plant before the next receipt of fresh fuel, this URIis considered closed, t

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E.8.5 (Closed) URI 60 293/94 26-01: Diesel Generator Turbo Assist Solenoid Valve-Testina Insoection Sqqng (92903)

The inspector reviewed the status of BECo's actions to address past problems with diesel generator turbo assist solenoid valves, Qbservations. Findinas, and Conclusions t On January 24,1994, the *B" emergency diesel generator (EDG) f ailed to stort within the l maximum allowable time interval due to the simultaneous f ailure of the two diesel turbo  ;

assist solenoid valves. The f ailure was determined to be caused by moisture in the air -

system. BECo's immediate corrective actions included a modification to allow on-line  ;

testing and replacement of the solenoid valves, along with increasing the solenold valve >

test frequency from every 18 months to quarterly. The licensee subsequently developed field revision notice (FRN) 97 04 47 to replace the EDG turbo air assist solenoid valves with a different type of valve that is better able tn operate in the existing environmen The inspector reviewed the operational history for EDG start times and the maintenance history for the turbo assist solenold valves for the past two years. Inspection revealed that the diesel start times have been satisfactory and there have been no f ailures of the dolenoid valves. The inspector also verified that BECo developed a FRN and issued a maintenance request to replace the subject solenoid valves. Implementation of the FRN is scheduled to be performed by December 1997. Based on the satisf actory performancu of the EDG and solenoid valves os er the past two years and the planned implementation of FRN 97 04 07, this item is cor.s1dered closad, t

IV. PIART_EPPORT R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Tour of radioloalcal controlled areas insoection Sqoce (71750)

During plant tours of the plant, the inspectors reviewed radiological boundaries, postings and adherence to radiation work permit instructions. Also, the general plant cleanliness was observed, Q.bservatiom and Findinas Overall, plant cleanliness and radiological controls were good with two excr mons. During the tour of the condensate jockey pump room, the inspector observed thr/t a temporary pump down rig was laying on the floor with approximately one gallon of vsater drained from the pump hose. The pump down rig and water was beside a roped off contaminated

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leak. The inspector questioned radiation protection (RP) personnel at the redline area whether or not the pump down rig was used to dewater the barrels within the contaminated area. The RP technician later informed the inspector that the rig was used i by waste control technicians to dewater the barrel. The RP technician determined that the pump down rig should have remained in the contaminated area or had the water drained from the rig and any water spillage from the rig hoses cleaned up. The standing water was wiped up. Problem report no. 97.3249 was initiated to document, evaluate and take i corrective actions. The inspector noted that the water was not contaminated and no personnel contamination resulte A second observation was made in the torus room where the cleanliness and housekeeping degraded slightly since the previous routine inspection pet;od. A protective clothing (PC)

receptacle was full and overflowed with several pieces of used PC laying on the floor. A lagging pad for a flanged mechanical pipe joint was removed and laying on the floor. In  ;

addition, some debris, including dirt, tie wraps and old tags was evident. The inspector informed the assistant operations department manager of these housekeeping issues im ,

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evaluation and corrective action.

Conclusions Two instances of poor radiological practices were observed including an overflowed protective clothing receptacle and spillage of water from a ' watering rig outside of a posted contamination area. Minor cleanliness issues were observed in the torus roo V. MANAGEMENT MEETINGS X1 Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inspection at an exit meeting conducted on December 8,1997. The licensee acknowledged the findings presented . The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

Additionally, a public Systematic Assessment Licensee Performance (SALP) meeting was held onsite on November 13,1997,in the Engineering Support Building.' Mr. Hubert Miller, NRC Region l Administrator, f acilitated the meeting which focused on overall station performance. Prior to the meeting on November 12,1997, Messrs. H. Miller, J. Wiggins, B.Boger, W. Lazarus, S. Chaudery, A. Wang, R. Conte (all from NRC) conducted plant tours led by the NRC Resident inspector staff and also interviewed plant workers and supervisory level personne Also, on November 5,1997, an emergency preparedness (EP) drill with full Commonwealth of Massachusetts participation was conducted at PNPS. An NRC inspection team led by Mr. J. Lusher, NRC Region l Senior EP Specialist, observed the EP drill to ensure compliance with EP regulations. The results of this inspection were documented in NRC

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Inspection Report No. 50 293/97 10. Lastly, a public FEMA /NRC exit meeting was held  !

on November 10,1997, at Memorial Halllocated in Plymouth, M i X4 - Review of UFSAR Commitments l A recent discovery of a licensee operating their f acility 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 practices and procedures. While performing inspections discussed in this report, inspectors reviewed the applicable portions of the UFSAR. No inconsistencies were note ,

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20 l 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 3~

IP 71001: Licensed Operator Requalification Program Evaluation IP 71707: - Plant Operations IP 71750: Plant Support Activities IP 82301: Evaluation of Exercises for Power Reactors IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor

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Facilities .f IP_92901: Followup Operations j

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IP 92902: Followup Maintenance lP 92903: Followup Engineering IP 92904: Follow ,p Plant Support IP 93702: Prompt Onsite Response to Events at Operating Power Reactors i

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i ITEMS OPENED, CLOSED, AND UPDATED i i

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50 293/97 11 01 . VIO Tagging procedure 1./ .5; two examples of inadequate tegou /97-11 02 !FI internal Nuclear ' / Concern from 1989 7

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50 293/97 11 03 UNR SSW Pump Motor haft Breakage Corrective Actions

Closed 50 293/97-09 LER RCIC system inoperable due to turbine over speed trip during surveillance

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50 293/95 11-01 LER RCIC system made inoperable i

50 293/95 06 LER RCIC system inoperable due to turbine steam problem during test 50 293/97 02-04 URI Compliance with 10 CFR 70.24 ,

50 293/96 80 01 IFl Power to flow man guidance 50 293/97 04 01 IFl Adequacy of TS re: offsite power

.50 293/97 00-01 IFl SSW motor shaft shear .

t 50 293/94 26-01 UNR EDG turbo assist solenoid valves Uodated None

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LIST OF ACRONYMS USED ALARA As Low As is Reasonably Achievable APRMs Average Power Range Monitors BECo Boston Edison Company CFR Code of Federal Regulations C3D Control Rod Drive CS Core Spray EP Emergency Preparedness EPIC Emergency and Plant Information Computer F8F Engineored Safety Feature gpm gallons pc' minute I&C Instru'.1entelon and Controls IFl Inspe tion Ft' low Up Item

'R Inspec' ion Re sort LER Licens% Ennt 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 Reactor Operator-TM Temporary Modification TS Technical Specification UFSAR Updated Final Safety Analysis Report WWM Work Week Manager