IR 05000293/1998006

Insp Rept 50-293/98-06 on 980609-0725.No Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML20239A130
Person / Time
Site:	Pilgrim
Issue date:	08/28/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20239A116	List: IR 05000293/1998006 ML20239A114 ML20239A123
References
50-293-98-06, 50-293-98-6, NUDOCS 9809080248
Download: ML20239A130 (26)

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U.S. NUCLEAR REGULATORY COMMISSION REGION I '

License No.: DPR-35 Report No.: 98-06 Docket No.: 50-293 Licensee: BEC Energy 800 Bo'ylston Street Boston, Massachusetts 02199 l Facility: Pilgrim Nuclear Power Station l Inspection Period: June 9,1998 - July 25,1998 Inspectors: R. Laura, Senior Resident inspector R. Arrighi, Resident inspector L. Peluso, Environmental Monitoring Specialist P. Cataldo, Millstone Unit 1 Resident inspector Approved by: Curtis J. Cowgill, lil, Chief, Projects Branch 5 Division of Reactor Projects i

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EXECUTIVE SUMMARY Pilgrim Nuclear Power Station

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NRC Integrated inspection Report 50-293/98-06 l

- This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers resident inspection for the period of June 9,1998, through July 25,1998. In addition, it includes the results of announced inspections by a regional radiological effluents specialist and a resident inspector from Millstone Unit . Ooerations

The conduct of operations was professional and safety-conscious. Operators responded effectively to two plant transients. in one instance, operators gained control of a recirculation pump which suddenly increased in speed due to a controller malfunction. This action took only six seconds and averted a potential plant transient or reactor scram. (Section 01.1)

Reactor engineers and operators closely monitored minor reactor power oscillations and initiated action to lower power to prevent exceeding the instantaneous reactor power limit. (Section 01.1)

• Review of operator overtime data revealed that the licensee maintains sufficient licensed operations personnel to maintain adequate shift coverage without routine heavy use of overtime. (Section 01.1)

• The licensed and non-licensed operators were knowledgeable of plant and equipment status during equipment rounds. (Section 01.2)

Maintenance

• Routine work and surveillance activities were well planned and executed. (Section M1.1)

• The chemistry department was slow to report the initial indication of water in the core spray pump motor bearing oil sample. Good coordination was noted between the various departments to investigate and promptly restore the core spray pump to service once this issue was identified. (Section M1.1)

• The core spray pump unavailability time was properly accounted for during the maintenance activity in accordance with the maintenance rule requirement (Section M1.1)

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. An effective questioning attitude by the engineering staff led to the identification of the inadequate implementation of the 4.16kV degraded voltage protection logi Engineering evaluations were performed that concluded the emergency AC power system was operable but outside design. No immediate corrective actions were necessary; procedures were already in place to prevent this scenario from occurring, and the delay in restoring core cooling after a DBA-LOCA was rietermined not to cause the peak clad temperature to exceed the temperature lir6.ws specified in 10CFR 46. Corrective actions are being developed that will be implemented during the next refueling outage, presently scheduled for April 23,1999, to modify the degraded grid voltage relay setting. (Section E2.1)

. The engineering staff closely followed the guidance contained in NRC Generic Letter 91-18, Revision 1, during the identification and resolution of three adverse design control conditions involving a potential containment leakage path, electrical hot shorts and use of incorrect seismic damping factors. (Sections E2.2, E8.6 and F1.1)

Plant Suocort

. An effective radiological environmental monitoring program was implemented and maintained in accordance with regulatory requirements. (Section R1.1)

. The licensee effectively maintained and implemented a meteorological monitoring program in accordance with regulatory requirements. (Section R1.2)

. OA audits and surveillance were thorough and of sufficient depth to assess performance and implementation of the REMP. (Section R7.1)

. The contract laboratory continued to implement excellent QA/QC programs for the REMP, and continued to provide effective validation of analytical results and the programs are capable of ensuring independent checks on the precision and accuracy of the measurements of radioactive material in environmental media. (Section R7.2)

. Two violations wue identified by the NRC that related to the fire protection program. The licensee subsequently took thorough corrective actions including performing detailed root cause evaluations, issuance of two related LERs, detailed problem extent review for other degraded fire barriers, and prompt resolution of hardware and testing deficiencies. (Section F1.1)

. An ALARA concern that could potentially result in higher cumulative radiation exposure for operators who perform turbine star dard inspections was effectively addressed by the licensee. (Section 01.2)

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TABLE OF CONTENTS EX ECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii Summ ary of Plant Status .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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I 1. O PERATI O N S . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 L ~ 01.1 General Comments (71707) ............................1 01.2 ' Plant Tour Accompaniment . . . . . . . . . . . . .. . . . . . . . . . . . . . . 2 II . M AINTEN ANC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 M1' Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 M 1.1. : General Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 ~

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M8. - Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . ... . . . . . . . . 4 M8.1 (Closed) VIO 50-2 9 3 /9 7-13-01 . . . . . . . . . . . . . . . . . . . . . . . . . 4 M8.2 (Closed) VIO 50-2 9 3/98-0 2-0 2 . . . . . . . . . . . . . . . . . . . . . . . . . 4 M8.3 (Closed) LER 50-293/97-14-00and -01 ................... 5 M8.4 (Closed) LER 50-293/97-12-00 and -01 ................... 5 M8.5 (Closed) LER 50 293/97-16-01 . . . . . . . . . . . . . . . . . . . . . . . . . 6 lll . ENG I NEERI NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 E2' Engineering Support of Facilities and Equipment .................. 6 E (Closed) LER 50-293/98-14 and LER 50-293/98-15 . . . . . . . . . . . 6 E2.2 (Closed) LER 50-293/92-01, Supplement 1 . . . . . . . . . . . . . . . . . 8 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 E8.1 (Closed) VIO 50-2 9 3/9 7- 13-0 2 . . . . . . . . . . . . . . . . . . . . . . . . . 9 E8.2 (Closed) VIO 50-2 9 3/98-01 -07 . . . . . . . . . . . . . . . . . . . . . . . . . 9 E8.3 (Closed) VIO 50-293/98-01 -08 . . . . . . . . . . . . . . . . . . . . . . . . 10 E8.4 (Closed) LER 50-2 9 3/9 7-2 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 E8.5 - (Closed) LER 50-293/97-23-00and -01 ..................10 j E8.6 . (Closed) LER 50-293/97-24 . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1

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IV. PLANT SUPPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 R1' Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 11 ,

R Implementation of the Radiological Environmental Monitoring Program ( R E M P) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 )

.R1.2 Meteorological Monitoring Program (MMP) ................ 13 i R6 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 14 R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 14 R7.1 Quality Assurance Audit Program ......................14 R7.2 Quality Assurance of Analytical Measurements . . . . . . . . . . . . . 15  ;

F1 Control of Fire Protection Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 16 F1.1 (Open) VIOs 98-06-08 and 09 (Closed) URI 98-05-01 . . . . . . . . 16 V. M AN AG EMENT M EETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

~X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 iv

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X3 Management Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 ITEMS OPENED, CLOSED, AND UPDATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 LIST OF ACRONYM S USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 v

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REPORT DETAILS Summarv of Plant Status Pilgrim Nuclear Power Station (PNPS) began the period operating at approximately 100 percent power. There were two power reductions during this inspection period. On July 16 operators lowered power when the "A" train third point feedwater heater isolated on high water level. Repairs were made in the condenser bay and power was returned to full power on July 17. On July 24 operators reduced power to approximately 50 percent for a planned thermal backwash of the main condenser to reduce biofouling. Also, main steam isolatior valve and control rod scram time testing was performed during the downpowe The plant was at 50 percent power at the end of the report perio l. OPERATIONS

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01 Conduct of Operations'

01.1 General Comments (717071 Using inspection Procedure 71707, the inspector conducted frequent reviews of ongoing plant operations. The inspector observed proper control room staffing, effective pre-evolution briefings, and plant behavior that was commensurate with the plant configuration and plant activities in progress. The inspector discussed anomalies noted during plant tours with the nuclear watch engineer who initiated corrective actions, as neede Operators properly responded to the loss of the third point feedwater heater and recirculation pump transients. Prompt operator identification and response was noted. In the case of the increasing recirculation pump speed, the 905 operator responded rapidly (within about six seconds) to lock-out the scoop tube. This response averted a plant transient. Later, operators followed plant procedures to manually lower the recirculation pump speed by adjusting the scoop tube position on the motor-generator set. The inspector interviewed the operator and reviewed the computer alarm printout and identified no problem l During this inspection period, the inspector performed a walkdown of accessible plant areas and verified proper positioning of type "A" containment isolation valves, in addition, a review of selected licensed and non-licensed operator (total of 15 operators) overtime sheets for the period between April and June 1998 was performed to ensure compliance with PNPS procedures and NRC Generic Letter 82-12. No problems were noted. The licensee maintains sufficient licensed operations personnel to maintain adequate shift coverage without routine heavy use of overtime.

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' 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 j address all outline topic l

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Operators and reactor engineers closely monitored the effects of small reactor )

power oscillations due to a phenomenon in boiling water reactors known as

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" bistable vortexing." The vortexing occurs in the reactor recirculation syste During this period, the licensee administratively lowered the reactor power limit to 1980 MWt as opposed the licensed limit of 1998 MWt. This power reduction was

' to ensure the oscillations did not result in a technical specification violation of exceeding the instantaneous limit of 102 percent reactor power. Fine tuning of the turbine control system was planned to mitigate the oscillation .2 Plant Tour Accompaniment Inspection Scope (71707)

The inspector observed licensee performance of operations department plant tours consisting of a reactor building tour, turbine building tour, and outside rounds (e.g., j diesel generator and intake structure). In addition, the inspector observed licensed operator performance in the control roo Observations and Findinos The plant tours were performed by licensed and non-licensed operators that were knowledgeable of plant equipment status. In addition, the licensed operators in the control room were well briefed on emergent issues and the subsequent effect on the plant. During the turbine building tour, the inspector identified an ALARA (as low as reasonably achievable) concern that the licensee immediately addresse An operator mariipulated a remote, robotic camera for general observations of the turbine standard in an area with a posted radiation field of 25 - 30 mrem /hr dose ,

rate. For approximately 10 minutes the operator manipulated the robotic camera in j the higher dose rate area while a majority of the turbine periphery was shielded to l keep dose rates to a minimum. The area that contained the controls for the robotic '

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camera had no permanent or portable shielding to minimize dose to the operato While the immediate ALARA impact was low, the inspector was concerned about

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the cumulative effect of the radiation dose over longer periods of tim The inspector informed the nuclear watch engineer (NWE) and the acting radiation protection supervisor. The inspector was briefed by the NWE on the expectations ]

of the operations department for operators who performed the turbine inspectio The acting radiation protection supervisor was not fully cognizant of the scope of )

the turbine inspections, and indicated that he would contact the operations l l department to evaluate this potential ALARA concern. At the end of the inspection l period, the operations department manager informed the inspector that the remote controls were relocated behind existing shielding to further reduce the cumulative dose.

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3 Conclusions The licensed and non-licensed operators were knowledgeable of plant and equipment status. An ALARA concern that could potentially result in higher cumulative radiation exposure for operators who perform turbine standard inspections was effectively addressed by the license . MAINTENANCE M1 Conduct of Maintenance M 1.1 General Maintenance Insoection Scooe (62707/61726)

The inspector observed portions of selected maintenance and surveillance activities to verify proper use of approved procedures, conformance to limiting conditions of operation, and correct system restoration following maintenance and/or testin Portions of the following activities were observed:

• 8.7. "MSIV Twice Weekly"

- 8.M.1-21 "RPS Channel Test Switch"

• 19801788 "A" Core Spray Pump Bearing Oil Flush and Change Out a 19702406 EDG Pipe Support

. 19801819 "A" Recirculation Motor-Generator Set Controller Observations and Findinas The inspector found that all observed work activities were well planned and executed. All work observed was performed with the work package present and in active use. The inspector verified that the applicable technical specification (TS)

action statement was entered upon disabling of equipmen On July 14,1998, an abnormal water content (trace amount) was detected in an oil sample from the "A" core spray (CS) pump lower bearing. The initial surveillance results were known about two weeks earlier, but the chemistry department was slow to provide the results to engineering for evaluation. Once this was identified, a proper evaluation was made and the issue reported to the operations departmen Upon such notification, a priority one work request was generated and the motor bearing re-sampled. The second sample indicated approximately 50 percent water through visual observation. The applicable TS action statement was entered and the CS pump declared inoperable. The lower CS pump motor bearing was flushed and changed out to remove any residual wate The licensee performed a final oil analysis and identified the total water content to have been approximately 1.2 percent water. The disparity in the sample flask water I content was attributed to sampling technique (the level from which the sample was

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drawn). An operability evaluation was performed that concluded the amount of water in the oil was not enough to affect the bearing condition or operating environment. The licensee was unable to identify the water source and attributed the cause of the water to have accumulated from the April 1998 upper bearing cooling coil leak (reference NRC Inspection Report No 50-293/98-02, Section M3.1).

The inspector monitored the CS pump oil flush and visually inspected the two sample oil flasks. There was no visible indication of water in the first sample flas A review of the prior inservice pump testing data sheet revealed no abnormal pump / motor readings. Geod coordination was noted between the engineering, chemistry, and maintenance departments upon identification of water in the "A" core spray pump lower bearing. The inspector noted that the licensee properly accounted for the core spray pump unavailability time in accordance with the maintenance rule requirement c. Conclusions Observed activities were well planned and executed. The chemistry department was slow to report the initial indication of water in the core spray pump motor bearing oil sample. Good coordination was noted between the various departments to investigate and promptly restore the core spray pump to service once this issue was identified. The core spray pump unavailability time was properly accounted for during the maintenance activity in accordance with the maintenance rul M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) VIO 50-293/97-13-01: Wrono ATWS Relav Installed The violation related to the installation of the incorrect ATWS system electrical relay which overheated and became damaged. The root cause of the violation was failure to follow work control process instructions. Specifically, the job was changed to

" task ready" when all parts were not yet available. Subsequent to the relay failure, the correct relay was installed and the ATWS system returned to operable. The longer-term corrective actions included a change to procedure 1.5.20, " Work Control Process," to enhance the " task ready" review process. Revision 11 was issued to procedure 1.5.20 in February 1998. Additionally, lessons learned meetings were conducted with maintenance, instrumentation and controls, and engineering personnel to review the events that led to the installation of the incorrect ATWS relay. This violation is close M8.2 (Closed) VIO 50-293/98-02-02: Inadeouate Work Packaae for Core Sorav Pumo The violation related to an inadequate work package that resulted in damage to an internal core spray pump motor cooling coil and increased radiation worker exposure. The damaged cooling coil was replaced. The longer-term corrective !

actions involved changes to design documents and work instructions, as needed, I and human error prevention techniques. This violation is close l

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M8.3 (Closed) LER 50-293/97-14-00and -01: Transmitters Not Calibrated in Accordance with Technical Specifications (TS)

This LER documented that several pressure transmitters had been calibrated with the plant on-line rather than during a refueling outage as required by TS. The calibration of the transmitters was scheduled during the April 1997 refueling outage but had been rescheduled to allow the performance of emergent work activitie The licensee had performed an evaluation cor cluding that the transmitters could be safely calibrated on-liae but incorrectly interpreted the TS surveillance requirement The transmitters were calibrated with satisfactory as-found result To prevent recurrence, the master surveillance tracking program (MSTP) nodes l associated with these instruments were revised to be consistent with TS. A review was performed of the MSTP nodes and TS tables to provide assurance of consistency and accuracy for devices that are coded for testing during refueling ;

outages. The inspector conducted an on-site review of the LER and verified that the '

MSTP nodes associated with the identified transmitters had been revised to indicate a refueling outage frequency. The inspector determined that the failure to perform ,

the calibration of the pressure transmitters during the cycle 11 refueling outage was '

a violation of technical specifications. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-293-98-06-01). This LER is close M8.4 (Closed) LER 50-293/97-12-00and -01: Failure of Relav to Eneraize Durina Surveillance Test This LER documented that the relay that initiates a portion of the pipe break detection circuit in the residual heat removal system channel "B" circuitry did not energize during a surveillance test. The relay did not energize because contacts 9-10 of relay 10A-K10B were incorrectly set. This incorrect setting was attributed to improper alignment during initial setup of the relay. The relay had been replaced in May 1995 and satisfactorily passed the post-maintenance tes '

Corrective actions included declaring the RHR low pressure coolant injection (LPCI)

mode inoperable and replacement of the relay. Additional corrective actions included revising the procedure used for testing HFA relays (3.M.3-30) and performing a visualinspection of HFA relays to determine the extent of conditio The inspector conducted an on-site review of the LER, verifying procedure 3.M.3-30 was in the process of being revised and a training module was implemented that provided hands-on experience for HFA relay installation and testing. The inspector determined that one channel (less than the minimum nurnber (2) of instrument channels) for initiation of the RHR/LPCI function was inoperable greater than the time specified in the limiting condition of operation in violation of technical i

specifications. This non-repetitive, licensee-identified and corrected violation is l being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-293-98-06-02). This LER is closed, f

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M8,5 (Closed) LER 50-293/97-16-01: Recirculation System Loon "B" Pumo Trio This LER was previously reviewed and documented in Section 08.6 of NRC l Inspection Report No. 50-293/98-01 as a Non-Cited Violation. This LER is close l11. ENGINEERING E2 Engineering Support of Facilities and Equipment E (Closed) LER 50-293/98-14 and LER 50-293/98-15: Emeraencv Diesel Generator Deoraded Voltaae Time Delav Relav Settina insoection Scone (37551)

On June 22,1998, the licensee made two NRC notifications pursuant to 10 CFR 50.72(b)(1)(ii)(B) for conditions involving the under-voltage protection on the unit auxiliary transformer power supply to the Class 1E buses. The inspector reviewed the problem identification and corrective action aspects of theses issue Observations and Findinos As part of the design basis information (DBI) program review of the emergency diesel generator (EDG), the licensee identified two conditions that were outside the design basis of the plant. Both issues deal with the 4.16kV degraded voltage time-delay relay setting being set non-conservatively. This condition can affect (1) the time for the EDG to power the 4.16kV safety-related buses and (2) result in the tripping of the core spray (CS) pump after a loss of coolant accident (LOCA) signa The degraded voltage time-delay relay was installed in 1982 as the result of a NRC required modification (reference NRC letter dated June 3,1977).

Issue (1) - During a design basis LOCA concurrent with a degraded grid voltage, restoration of continuous power to the emergency buses (A5 and A6) via the EDG would exceed the time stated in the Updated Final Safety Analysis Report (UFSAR)

accident analysis. The UFSAR analysis assumes a total time of 14.4 seconds for the EDG to restore power to the emergency buses. The degraded voltage time-delay relays are set to trip the breakers connecting the startup transformer (preferred AC power source) to safety-related buses A5 and A6 after a time delay of 10.24 sec. +/- 0.36 seconds. The diesel generator output breaker will not close and restore power to the safety-related buses until after the diesel time delay permissive relay times out about 10.2 seconds after it senses no power to the startup transformer. This delay exists even though the diesels start due to the LOCA signal several seconds before. Thus, the total delay time (between detection of degraded voltage, disconnection from offsite power, and restoration of power by l the EDG to safety-related buses A5 and A6) of 20.8 seconds exceeds the 1 seconds assumed in the UFSAR analysis. This results in a delay of the core standby cooling systems (core spray and low pressure coolant injection pumps) injecting into the core and results in a small increase in the peak clad temperature.

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issue (2) - The degraded voltage relay is designed such that it shall not result in a failure of safety systems or components. The licensee identified that a degraded voltage condition (within a specific band) coincident with a LOCA could cause the coro spray (CS) pump motors to trip on over-current prior to the degraded voltage protection isolating buses A5 and A6. Manual operator action would be required to restart the CS pumps, contrary to that assumed in the accident analysi The licensee's operability evaluation addressing the discrepancy in the timing of restoring emergency power (issue 1) concluded that the emergency AC power systems were operable and the safety-related systems powered by buses A5 and A6 would meet their design basis functions. This evaluation considered the two most limiting, single-failures considered in the LOCA analysis. Based on this review, the licensee concluded for the limiting case that the peak clad temperature would increase approximately 15 degrees F (to 1836 degrees F), which is less than the design basis temperature of 2200 degrees The operability evaluation to address the tripping of the CS pump (issue 2) was based on interim steps to prevent the scenario from occurring. Procedures are in place to prevent placing the plant in a degraded voltage condition. The licensee's bulk power dispatcher, REMVEC, has procedures in place to notify Pilgrim in the event that they cannot ensure adequate grid voltage if ine Pilgrim unit should tri The REMVEC computer alarms if a system configuration develops which does not support the PNPS requirements. PNPS procedure 2.4.14.4 requires the EDG be started and placed on the emergency buses if notified by REMVEC that the 345kV feed to Pilgrim cannot be maintained above 340kV, or if the degraded voltage alarm annunciated in the control room. A standing order requiring operators to notify REMVEC once per shift to verify adequate grid voltage was implemented until final corrective actions are completed. Additionally, in the event that a core spray pump trips as a result of a degraded voltage condition, manual operator action would be required to start the CS pump. The alarm response procedure for the CS pump trip requires that the CS pump controls be taken to trip and then restarted (this is accomplished in the control room). The inspector reviewed the referenced procedures and verified that they contained steps to mitigate this condition. Long-term corrective actions are scheduled to be implemented during the cycle 12 refueling outage, scheduled for April 23,199 The inadequate implementation of the 4.16kV degraded voltage protection logic was a violation of design control requirements contained in 10CFR 50, Appendix B, Criterion 111. However, this item was licensee-identified during its validation of the emergency diesel generator design basis document initiative. The immediate and long-term corrective actions were comprehensive and either completed or appropriately scheduled for completion within a reasonable time. The issue also was not likely to be identified by routine licensee activities. Therefore, in accordance with the NRC's Enforcement Policy Vll.B.3, involving old design issues, this violation was not cited. (NCV 50-293/98-06-03). LERs 50-293/9814and 98-15 are close (.

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8 Conclusion The inspector noted that an excellent questioning attitude by the engineering staff led to the identification of the inadequate implementation of the 4.16kV degraded voltage protection logic. Engineering evaluations were performed that concluded the emergency AC power system remained operable. This old design issue was a violation of 10 CFR 50 Appendix B Criterion til but was not cited in accordance with Enforcement Policy Vil.B.3. Corrective actions to modify the degraded grid voltage relay settmg are being developed that will be implemented during the next refueling outage, presently scheduled for April 23,199 E2.2 (Closed) LER 50-293/92-01 Supolement 1. Class l Pioina Seismic Damoino Ratios The inspector reviewed Supplement 1 to LER 92-01, " Class l Piping Seismic Damping Ratios," which was issued as a voluntary LER during this inspection period. LER 92-01, dated February 21,1992, reported a condition that involved the incorrect use of seismic damping factors for some Class I piping. LER 92-01 was previously reviewed by the NRC in section 7.2.1 of NRC Inspection Report No. 50-293/92-04. The 1992 operability evaluation addressing this issue concluded that the piping remained operable. LER 92-01 also reported that the damping ratios were increased from the values specified in the UFSAR without NRC review and approval. Licensee safety evaluations (SE) 1421, dated May 24,1982, and SE 1697, dated August 23,1984, incorrectly determined that the changes did not involve an unreviewed safety question (USO). This deficiency was identified by the licensee during an operating experience revie The licensee performed a detailed engineering evaluation (EE) in October 1993 of the effects of using various damping values in the seismic analysis. Prior to restart from the cycle 11 refueling outage (RFO11), dated 2/97 - 4/97, the licensee recognized that the October 1993 EE needed to be submitted as an USQ to the NRC for review and approval. The NRR staff evaluated the October 1993 EE and determined that the licensee inappropriately developed an alternate method of deriving seismic inputs for the analysis of the aforementioned piping system Specifically, the use of a NRC Regulatory Guide 1.60 spectrum in the free field at the ground surface was not acceptable since the original design basis introduced the control motion at the building foundation level. The licensee initiated problem report (PR) 97.1026 to document, evaluate, and correct these issue Subsequently, the licensee performed design basis analyses for the affected piping systems to qualify the piping to the correct design criteria. The licensee determined that the previously unanalyzed piping met the lower, more conservative damping values required by the original UFSAR. As a result, the amendment request submitted to the NRC was withdrawn prior to the restart from RFO11. The licensee determined this was not reportable pursuant to 10CFR 50.72/73 criteria since the subject piping was qualifiable to the older and more stringent design criteria. The issue was no longer considered an RFO11 restart issu _ _ - - _ - - - - - ._

9 A licensee root cause evaluation was performed as part of the actions required by i PR 97.102 and identified two causes for the problem. First, the original safety evaluations (i.e.,1421 and 1697) failed to recognize that NRC review and approval j was required. Secondly, the open operability evaluation performed in 1992 was not '

tracked as a plant start-up restrain Licensee safety evaluations 1421 and 1697 did not provide sufficient basis to j determine that USQs did not exist. This was a violation of 10 CFR 50.5 However, the design basis analysis performed prior to restart from RFO11  !

determined that the subject piping was qualified using the original damping values :

listed in the UFSAR. This item was licensee-identified during its review of the licensing and design basis in support of a license change submittal. The corrective actions were comprehensive and performed within a reasonable time. The issue also was not likely to be identified by routine licensee activities. Therefore, in accordance with the NRC's Enforcement Policy Vll.B.3, involving old design issues, j this violation was not cited. (NCV 50-293/98-06-04) i E8 Miscellaneous Engineering issues (92903)

E (Closed) VIO 50-293/97-13-02: Reactor Vessel Flanae Metal Temperature Indication The violation related to repetitive problems with reactor vessel temperature j detectors which became an operational burden. The root cause was the lack of ownership to resolve temperature detector deficiencies. Engineering analysis determined that the problems experienced with the temperature detectors did not result in exceeding the code stress allowable for the reactor vessel. The licensee determined that the 145 degree F differential temperature limit was a hold over requirement from initial plant testing and was not necessary for current plant conditions. Subsequently, the licensee requested a license amendment to delete the requirement to monitor the 145 degree F limit for reactor vessel flange to shell ;

differential temperature. The NRC reviewed and approved License Amendment l No.175, effective on June 19,1998, that deleted this requirement. The inspectors l found that the licensee's actions were reasonable and complete. This violation is close E8.2 (Closed) VIO 50-293/98-01-07: Late 50.72 Notification The violation related to the late NRC notification of a design deficiency involving the emergency diesel generator fuel oil storage system. The cause for the late l

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notification was human error. An engineer had prepared an engineering evaluation l

but did not initiate a problem report that triggers a deportability review. Engineering management conducted training to emphasize the importance of following the problem reporting process. Engineers were reminded that a degraded condition may not render a system inoperable; however, the condition may still be reportable as outside the design basis. The inspectors found that the licensee's actions were reasonable and complete. This violation is closed.

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E8.3 (Closed) VIO 50-293/98-01-08: UFSAR Uodate Process l

The violation was related to the untimely update of the Pilgrim UFSAR due to a program deficiency. The root cause was the UFSAR update procedure did not meet the intent of 10 CFR 50.71(e). The inspector verified that appropriate UFSAR program procedure changes were made. One such change required updating the UFSAR during the operational turnover phase of a design change rather than waiting until the modification close-out phase. The backlog of UFSAR change requests was updated in the October 1997 submittal to the UFSAR. The inspectors found that the licensee's actions were reasonable and complete. This violation is close E8.4 (Closed) LER 50-293/97-22: Temocrarv Power Cabies and Extension Cords Found Draoed or Wracoed to Class 1E Conduits This LER documented that temporary power cables and extension cords were found draped or tie-wrapped to class 1E conduits in violation of electrical separation criteria. This issue was prompted by a nuclear network message involving a similar problem at another power plant. The as-found cables / extension cords were provided with circuit protection and were tied to, or near, only one train of safety-related conduit The licensee promptly corrected this condition by removing, rerouting, or de-energizing the identified cables. Procedure 3.M.3-36, " Control of Temporary Power," was revised to incorporate the required separation criteria. The inspector conducted an on-site review of the LER and performed a plant walk-down that identified no other examples of this problem. The inspector also verified that procedure 3.M.3-36 was scheduled to be revised to include new electrical separation criteria guidance. The failure to adequately maintain the design of the plant (physical separation) was a violation of 10 CFR 50, Appendix B, Criterion Ill,

" Design Control." This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-293/98-06-05). This LER is close E8.5 (Closed) LER 50-293/97-23-00 and -01: Radwaste Buildina Truck Lock Door in a Configuration Inconsistent with Tornado Deoressurization Analvsis Assumption This LER documented that the radwaste building was in a configuration not consistent with the tornado depressurization analysis. The tornado analysis assumes a single door assembly. The licensee identified that two complete door assemblies were installed with both the inner and outer doors closed. Investigation revealed that the a second outer door was added due to the original door sticking in the open position. The original door was subsequently fixed; however, the replacement door was not removed. This condition existed for about three month Immediate corrective action was taken to open the outer door. The second door was subsequently removed. To prevent recurrence, the licensee developed a program to identify and control compartment barriers. The inspector conducted an

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on-site review of the LER and visually verified the radwaste door configuration and verified that the radwaste building door was included in procedure 8.C.42, "Sub-compartment Barrier Control Surveillance." The failure to adequately maintain the design of the plant is a violation of 10 CFR 50, Appendix B, Criterion Ill, " Design Control." This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-293/98-06-06). This LER is close E8.6 (Closed) LER 50-293/97-24: Configuration for Inertino the Primarv Containment was Outside the Desian of the Plant This LER documented that a potential bypass flow path existed through which the  ;

drywell air space could communicate with the torus air space. This condition existed only during swap-over from torus to drywell inerting when manually operated valve 9-HO-117 and the drywell and torus purge valves are open. The plant is placed in this valve 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 startup. This condition was identified during followup to a General Electric 10CFR 21 notification. This issue was .

previously discussed in NRC Inspection Report 50-293/97-11, Section E ,

As immediate corrective actions, the Operations Department Manager generated a tracking license condition of operation to prevent placing the plant in this configuration. PNPS procedure 2.2.70, " Primary Containment Atmosphere Control System," was subsequently revised to preclude placing the plant in this configuration. The inspector conducted an on-site review of the LER and verified that procedure 2.2.70 was revised to secure the torus lineup prior to inerting the  !

drywell. The failure to adequately maintain the design of the plant is a violation of design control requirements contained in 10CFR 50, Appendix B, Criterion Ill. This issue was licensee identified through the licensee's operating experience review; the  ;

corrective actions were comprehensive and performed within a reasonable time; and the issue was not likely to be identified as a result of routine licensee activitie Therefore, in accordance with the NRC's Enforcement Policy Vll.B.3, involving old design issues, this violation was not cited. (NCV 50-293/98-06-07). This LER is close IV. PLANT SUPPORT I

R1 Radiological Protection and Chemistry (RP&C) Controls j R Implementation of the Radiological Environmental Monitorina Proaram (REMP) Insoection Scoce (84750)

i The following areas of the REMP were assessed and reviewed: (1) selected sampling and analysis procedures; (2) analytical data from 1998; (3) selected sampling i techniques; (4) operability and calibration of air samplers; (5) discrepancy report logs; (6) Land Use Census results; (7) the Annual Radiological Environmental

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! 12 Monitoring Reports for 1996 and 1997; and (8) the Dose Assessment Reports to j the General Public for 1996 and 199 b. Observations and Findinas The REMP procedures provided appropriate guidance to prepare, sample, and analyze the environmental sample media. Sampling techniques, performed by -

Engineering Services (BECo's Watertown Laboratory personnel), were appropriate to collect environmental sample media. The air sampling equipment and water compositors were operable during 1996 to 1998 as evidenced by the sample logs and sample analysis results. The air sampling equipment calibration results were within the established tolerances, and calibrations were performed within the frequency specified in the procedure. All discrepancies in the REMP are reported 1 and logged to provide a means to execute immediate corrective actions, and to provide the input to the Annual Radiological Environmental Monitoring Repor Corrective actions for the discrepancies were immediate and were documented in the annual REMP repor The annual Land Use Census was performed in 1996 and in 1997, during the growing season, as required by the Technical Specifications (TS). The licensee conducted a thorough land use survey, including a resident, garden, and milk animal census, and also collected broadleaf vegetation. (Collection of broadleaf vegetation may be performed in lieu of a garden census for ground release plants according to the Branch Technical Position, "An Acceptable Radiological Environmental Monitoring Program," Revision 1, November 1979.) No significant changes were made to the REMP prograin as a result of the census. As a result of a previous garden census performed in 1992, the licensee determined that the highest and second highest deposition factors (D/Q) from the station vent and main stack, respectively, were in close proximity to the nearest garden location and the site boundary. At that time, the licensee added the four locations to the sampling program to augment the REMP. These locations provide the most accurate " worst case" dose assessment results. The licensee continues to collect broadleaf vegetation from these locations, in addition to the land use census, and in addition to the routine REMP locations to maintain a historical baseline of the environment around the Pilgrim statio The Annual Radiological Environmental Monitoring Reports included results of the environmental monitoring program, program changes, land use census, and interlaboratory comparison program, as required by TS. The reports providad a comprehensive summary of the results of the REMP around the site and met Section 6.9.C.2 of the TS reporting requirements. The Annual Dose Assessment Reports to the General Public contained a detailed assessment of the radiological impact on the public. Based on the dose assessment results, there was no significant radiological impact on the general public from plant operations.

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13 Conclusion

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' The licensee effectively maintained and implemented a radiological environmental monitoring program in accordance with regulatory requirement R1.2 Meteorological Monitorina Proaram (MMP) Insoection Scope (84750)

The following areas of the MMP were assessed and reviewed: (1) calibration procedures; (2) calibration methods and results; (3) site operations logs and modification records; and (4) functional check Observations and Findinas Under the oversight of Emergency Preparedness, contract personnel from Engineering Services are responsible for performing weekly surveillance, maintenance, and quarterly calibrations. Execution of surveillance, calibrations, biweekly exchanges of the strip chart recorder paper, and any occurrences pertaining to the primary and secondary towers were documented in the site operations log book and in the meteorological database modification recor Occurrences regarding any discrepancies, including transmitter (sensor) or translator malfunction were corrected immediatel The primary meteorological tower is equipped with wind speed, wind direction, and temperature sensors at the 33-foot and 220-foot elevations, and the 160-foot backup tower is equipped with wind speed, direction, and temperature sensors at the 33-foot and 160-foot levels. The meteorological data were available at the equipment house via digital display using a portable computer, at the control room via digital display from the system computer and via analog strip chart recorder The chart recorders in the control room and the instrumentation at the primary and backup towers were well maintaine The performance of implementing meteorologicalinstrumentation calibrations was reviewed. Channel calibrations were performed quarterly according to Regulatory Guide 1.23, " Meteorological Programs in Support of Nuclear Power Plants," and the Emergency Preparedness Administrative Procedure Manual, EP-AD-421,

" Surveillance, Maintenance, and Calibration of MEDAP Equipment." The licensee performs electronic alignments for the wind speed, wind direction, and temperature once per week and as part of the quarterly calibration, according to the calibration procedure. The licensee submits the wind speed and wind direction sensors together with the associated cupwheel and vane to a wind tunnel test facility, where the paired sets are subjected to wind tunnel tests including, a test to verify (1) the sensors respond accurately to a known speed and direction and (2) the starting threshold (wind speed). The output of the wind speed sensors are compared to a NIST traceable standard for three speeds, according to the ASTM D

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5096-96," Standard Test Method for Determining the Performance of a Cup

! Anemometer or Propeller Anemometer," July 1996. During the quarterly

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calibration, the licensee exchanges all the sensors and verifies the accuracy of the system for the wind direction and temperature channels upon installatio The accuracy of the wind speed sensors was not verified upon installation. This matter was discussed and reviewed against ASTM D 4480-93," Standard Test Method for Measuring Surface Wind by Means of Wind Vanes and Rotating Anemometers," April 1993 with respect to the standard method for verifying the accuracy of the entire wind speed channel. The licensee stated that the above ASTM would be reviewed for applicability to the affected calibration procedur The licensee planned to incorporate this method during the next calibration (August 1998) for the primary tower. The primary tower contains a mechanism to lower the {

sensors to ground level. The licensee will incorporate this method for the backup

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tower after examining the safety implications of performing the method on the towe Conclusion The licensee effectively maintained and implemented a meteorological monitoring program in accordance with regulatory requirement R6 RP&C Organization and Administration The organization and the responsibilities relative to oversight of the REMP and MMP was reviewed. The Chemistry Department has the responsibility to implement the Radiological Environmental Monitoring Program. The Facilities and Equipment Team of the Emergency Preparedness Department has the responsibility to ensure that surveillance, calibrations, and maintenance of the meteorological monitoring equipment was performed. There were no major changes in the organization and responsibilities pertaining to oversight of the REMP and MMP since the previous inspection conducted in October 199 R7 Quality Assurance in RP&C Act!vities R7.1 Quality Assurance Audit Proaram Insoection Scone (84750)

The Quality Assurance (QA) Oversight Program Review 97-03 and the supporting surveillance reports of the REMP were reviewed against criteria required by TS, the QA Plan, and the QA procedure Observations and Conclusions The Quality Assurance Group (QAG) performed surveillance of the REMP, including the meteorological monitoring program. The surveillance were the supporting documents for the 1997 Oversite Program Review. Seven surveillance were performed from January through October 1997. The surveillance were detailed in scope and effectively assessed the performance of the chemistry department. The l

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assessor's audit performance was good, in that specific REMP activities were directly observed and timely feedback regarding performance of the activity was provided and discussed. The Oversight Program review was impartial and summarized the surveillance to provide an overview of the program strengths and weaknesses. The QAG met the requirements of the QA Plan, Conclusion The licensee met the QA audit requirements of the " Quality Assurance Plan." The surveillance were thorough and of sufficient depth to assess performance and implementation of the REM R7.2 Quality Assurance of Analytical Measurements Inspection Scope (84750)

The inspector reviewed the following aspects of the QA/QC orogram of the primary contractor laboratory:

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the QA program (internal audits)

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the results of QC program (split, duplicate, blind samples);

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the results of the Interlaboratory Comparison (cross-check) program; and

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the semiannual QA reports, Observations and Findinos The QA/QC program for analyses of REMP samples is conducted by the primary analytical contract laboratory, Duke Engineering & Services Environmental Laboratory (DESEL), formerly Yankee Atomic Environmental Laboratory (YAEL). A tour of the laboratory was conducted during an inspection of the Millstone Nuclear Power Station, Units 1,2,&3,in April 1998. (NRC Integrated inspection Report No(s) 50-245:336:423/98-207 pertains). The laboratory continued the programs as usua The QA officer at the laboratory conducted independent audits of laboratory operations. The audits were methodical and provided very good insight for improvement where needed. The laboratory published a Quality Assurance report semi-annuall The laboratory has interlaboratory and intralaboratory QC programs and published the results in the semiannual QA reports. The intrataboratory QC program consisted of measurements of blind, duplicate, spike, and split samples. The results were within the acceptance criteria established in the laboratory's analysic procedure The laboratory continued to participate in the EPA Cross-Check Program (for drinking water) and the interlaboratory Comparison Program provided by a vendor (Analytics, Inc.), and several laboratories (for example, Department of Energy (DOE)

RESL and DOE EML). The quality control of environmental radioanalyses at DESEL

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emulated the internal process control program of National Institute of Standards and Technology (NIST). DESEL's participation in these programs was excellen Conclusion The contract laboratory continued to implement excellent QA/QC programs for the REMP; continued to provide effective validation of analytical results; and the programs were found capable of ensuring independent checks on the precision and accuracy of the measurements of radioactive materialin environmental medi F1 Control of Fire Protection Activities F1.1 (Open) VIOs 98-06-08 and 09: Defect in Accendix R Electrical Raceway Enclosure and inadeauate Corrective Actions From 1996 Quality Assurance Audit. (Closed)

URI 98-05-01: Fire Protection Deficiencies Inspection Scooe (71750.3'7551)

An additionalinspection was conducted to assess the significance of several fire protection deficiencies that were identified and de:umented in the previous routine inspection report. NRC inspection Report 50-293/98-05 opened URI 98-05-01, which documented three concerns involving a defect in an Appendix R raceway enclosure, lack of qualification testing, and inadequate corrective actions. During this inspection period, the inspector reviewed the licensee's corrective actions to resolve the identified deficiencies. The related LERs were also reviewed, LER 98-12, " Incomplete Installation Of Fire Barrier" and LER 98-13," Inconclusive Fire Barrier Enclosure Test Data." Also, LER 97-29 was reviewed which related to an electrical hot short issu Observations and Findinas The first issue documented as part of URI 98-05-01 related to the NRC identification of a defect on fire barrier 194.5018-encl.3. This defect was an irregularly shaped hole in the protective fire coating of the barrier that was about two square feet in area. The location of the defect was on the lower, backside of the barrier, which was not readily detectable. The defect, however, was of sufficient size to permit the protected cables to be exposed to the effects of any exposure fire in this fire area. The licensee's fire surveillance procedures defined this defect as one that prevented the barrier from meeting its acceptance criteria, and resulted in it being inoperable. As subsequently documented in LER 98-12, the licensee determined the cause of this defect was inadequate implementation of the original plant design change that installed the enclosure. This defect was considered a nonconforming condition that was outside the design basis of the plant. The defect existed since l 1980. The inspector determined that the existence of the defect rendered the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire enclosure inoperable and was a violation (VIO 98-06-08)of Section Ill.G.2 of Appendix R.

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The inspector reviewed the safety significance of this defect in the fire barrier. This i barrier contained "A" train safe-shutdown power cables that ran through the cable {

spreading room (CSR). The CSR has fire detection and Halon suppression capability, in the case that a fire would occur in the area. The licensee has established administrative controls for transient combustible materials and ignition sources for this area. In addition, during routine tours the resident inspectors have observed that the licensee's administrative controls and practices have been implemented well. The Pilgrim fire protection design also includes an alternate safe-shutdown system located at remote stations in the plant that was designed to operate independent of a fire in the CSR. Based on these factors, the inspector i determined that the potential safety consequence of the degraded barrier was lo !

l Also, the inspector reviewed LER 98-12 onsite and determined that it was of adequate quality and met the reporting requirements of 10 CFR 50.'/3. LER 98-12 is close The licensee had previously identified a fire protection issue relating to electrical hot shorts as reported in LER 97 29. Specifically, a fire in the CSR could result in an electrical hot short causing spurious operation and damage to the RHR shutdown cooling suction valves. Such fire-induced damage could impair the capability to shut down the plant and maintain it in a safe shutdown condition. The cabling for the two RHR suction valves runs through the CSR where fire barrier 194.5018-enc is also located. The licensee implemented a temporary modification to reduce the potential for hot shorts for the RHR suction valve LER 97-29 was reviewed and is closed. The hot shorts issue was a violation of Appendix R Section Ill. G. The licensee has identified and acknowledged the violation in the LER and is taking appropriate compensatory and comprehensive corrective measures within a reasonable time. In accordance with Section Vll. of the Enforcement Policy, the NRC is exercising discretion with respect to this violation of 10CFR50, Appendix R requirements. (NCV 50-293/98-0610)

The second and third issues documented as part of URI 98-05-01 related to the lack of configuration specific qualification testing for three raceway enclosures and inadequate corrective actions to resolve the findings from a 1996 triennial fire protection audit (QA Audit 96-04). Previously, the inspector identified several concerns with the evaluation and close-out of findings from Audit 96-04. The most significant issue involved the lack of configuration specific qualification testing for three Appendix R electrical raceway enclosures which used Pyrocrete fireproofin The inspector determined that the configuration specific testing concerns raised in Audit 96-04 were valid and that the line organization failed to implement adequate corrective action. This was the first example of a fire protection program corrective action violation. Subsequently, the licensee issued LER 9813 which reported the l- lack of configuration specific testing as a potentially nonconforming condition outside the riesign basis of the plan Other examples of inadequate corrective actions related to Audit 96-04 were also identified. The second example of inadequate corrective actions relates to a gap i

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j 18 I between a wall and fire barrier 194.504C-enct 2 that was identified during the ,

audit, but was not corrected. A third example of inadequate corrective action dealt .!

with three fire protection engineering evaluations (FPEEs), which had not been evaluated pursuant to station requirements. Failure to promptly correct identified l conditions adverse to quality was a violation of the Pilgrim License Conditions 3F, !

the Pilgrim Fire Protection Program and the Pilgrim Quality Assurance Program . (VIO 98 06-09)

Two of the three electrical raceway enclosures located in the lower switchgear room were recently removed to install newly designed and approved fire enclosures due to the lack of configuration specific testing. The third enclosure located in the CSR, which had contained the earlier described defect, remained intact and was repaired. At the end of the inspection period, the licensee was evaluating the continued need for the 3-hour rating because the potential for any hot shorts was resolved by a modification. The inspector determined that the licensee took prompt corrective action to resolve the identified hardware and testing deficiencie !

A licensee root cause evaluation determined that the inadequate resolution of the i Audit 96-04 findings resulted from human errors. Time-pressure and consolidatin I two problem reports into one contributed to these errors. Several other contributing i causes were also identified. For example, although Audit 96-04 identified several important issues, QA did not effectively follow-up during the next annual fire j program audit in 1997. Also, a lack of critical self-assessment was evident by the j fire protection engineering line organization. The inspector determined that the root

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cause evaluations were detailed and provided valuable, critical self-assessmen . The inspector reviewed other fire protection audits conducted by the QA staff and did not identify similar problems. The inspector determined the corrective action deficiencies were isolated to the 96-04 audit. Also, as part of the initial problem extent determination related to degraded fire barriers, the licensee initiated a random sampling inspection. The licensee inspected 2495 fire barrier penetrations from a total population of 5264 and found only one other significant fire barrier degradation resulting in the barrier being inoperable, The degradation in this barrier resulted from corrosion of a conduit which was located in the intake structure. The licensee

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judged that the problems with the three aforementioned electrical raceway enclosures constructed with Pyrocrete fireproofing material were atypical and that overall fire penetrations were in good condition. The inspector determined that the licensee conducted a thorough extent-of-condition revie '

. Conclusions  ;

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' Two violations were idsntified by the NRC related to the fire protection program implementation that were caused by poor human performance in determining the

- scope of identified problems and in tracking the completion of corrective action effectiveness. The inspector concluded that this problem appeared isolated. The licensee's current corrective actions were thorough, including detailed root cause evaluations, issuance of two related LERs, detailed extent-of-condition reviews for

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V. MANAGEMENT MEETINGS

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X1 Exit Meeting Summary The inspector presented the findings of this inspection to members of the licensee management on August 11,1998. The licensee acknowledged the findings presente X3 Management Meeting Summary l On June 29,1998, Mr. William Axelson, Region 1 Deputy Regional Administrator, Mr. Curtis Cowgill, NRC Region 1 Projects Branch Chief, and Mr. Eugene Kelly, NRC Region 1 Systems Engineering Branch Chief, visited the site for a facility tour, interviewed members of the licensee workers and managers, and discussed current inspection issues with the NRC resident inspector staf .

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20 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 Operations IP 71750: Plant Support Activities IP 82301: Evaluation of Exercises for Power Reactors IP 84750-02 Radioactive Waste Treatment, and Effluent and Environmental Monitoring l lP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901: Followup - Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering .

IP 92904: Followup - Plant Support  !

IP 93702: Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND UPDATED l Opened j VIO 50-293/98-06-08 Defect in Appendix R Electrical Raceway Enclosure j VIO 50-293/98-06-09 Inadequate Corrective Actions From 1996 Quality Assurance Audit Closed LER 50-293/92-01-01 Class 1 Piping Seismic Damping Ratios LER 50-293/97-14-00/01 Transmitters Not Calibrated in Accordance with Technical Specifications (TS)

LER 50-293/97-12-00/01 Failure of Relay to Energize During Surveillance Test LER 50-293/97-16-01 Recirculation System Loop "B" Pump Trip i

LER 50-293/97-22 Temporary Power Cables and Extension Cords Found Draped l or Wrapped to Class 1E Conduits J LER 50-293/97-23-00/01 Radwaste 8uilding Truck Lock Door in a Configuration inconsistent with Tornado Depressurization Analysis Assumption I

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LER 50-293/97-24 Configuration for inerting the Primary Containment was Outside the Design of the Plant LER 50-293/97-29 SDC Suction Valves Vulnerable to Damage from Potential Failure LER 50-293/98-12 Incomplete installation Of Fire Barrier LER 50-293/98-13 inconclusive Fire Barrier Enclosure Test Data LER 50-293/98-14 Emergency Diesel Generator Degraded Voltage Time Delay

& 98-15 Relay Setting NCV 50-293/98-06-01 Transmitters Not Calibrated in Accordance with the Tech Specs NCV 50-293/98-06-02 Failure of Relay to Energize During Surveillance Test NCV 50-293/98-06-03 Degraded Voltage Restoration Time not Consistent with Safety Analysis Assumptions and Non-Conservative Degraded Voltage Trip Setpoint NCV 50-293/98-06-04 Class 1 Piping Seismic Ratios NCV 50-293/98-06-05 Temp Power Cables and Extension Cords found Draped to Class 1E Conduits NCV 50-293/98-06-06 Both Radwaste Building Trucklock Doors Closed NCV 50-293/98-06-07 Plant Valve Configuration for inerting the Primary Containment NCV 50-293/98-06-10 SDC Suction Valves Vulnerable to Damage From Potential Failure URI 50-293/98-05-01 Fire Protection Deficiencies VIO 50-293/97-13-01 Wrong ATWS Relay Installed due to Work Control Error VIO 50-293/97-13-02 Reactor Vessel Flange Metal temperature Indication VIO 50-293/98-01-07 Late 50.72 Notification VIO 50-293/98-01-08 Untimely UFSAR Update Process VIO 50-293/98-02-02 Inadequate Work Package For Core Spray Pump Motor I

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, LIST OF ACRONYMS USED

'ALARA As Low As is Reasonably Achievable BECo Boston Edison Compan CFR _ Code of Federal Regulations CS Core Spray EP' Emergency Preparedness I&C Instrumentation and Controls IFl Inspection Follow-Up item IR Inspection Report

'LER Licensee Event Report MMP Meteorological Monitoring Program NCV Non-Cited Violation NIST __ National Institute of Standards and Technology 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 QA Quality Assurance QC Quality Control-REMP Radiological Environmental Monitoring Program RHR Residual Heat Removal RP Radiological Protection SAL Systematic Assessment of Licensee Performance

'TS Technical Specification UFSAR Updated Final Safety Analysis Report VIO Violation

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