IR 05000293/2007005

IR 05000293-07-005 on 10/01/2007 - 12/31/2007 for Pilgrim Event Follow-up

ML080320209
Person / Time
Site:	Pilgrim
Issue date:	02/01/2008
From:	Powell R J NRC/RGN-I/DRP/PB5
To:	Bronson K Entergy Nuclear Operations
Powell R J, RI/DRP/610-337-6967
References
IR-07-005
Download: ML080320209 (33)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406

February 1, 2008

Mr. Kevin Bronson Site Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508

SUBJECT: PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/2007005

Dear Mr. Bronson:

On December 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed report documents the results, which were discussed on January 9, 2008, with you and members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents one self-revealing finding of very low safety significance (Green) for which no violation of NRC requirements was identified.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be available electronically for public ins pection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ Raymond J. Powell, Chief Projects Branch 5 Division of Reactor Projects Docket No. 50-293 License No. DPR-35

Enclosure:

Inspection Report 05000293/2007005

w/Attachment:

Supplemental Information

cc w/encl:

SUMMARY OF FINDINGS

IR 05000293/2007-005; 10/01/2007-12/31/2007; Pilgrim Nuclear Power Station; Event

Follow-up.

The report covered a 13-week period of inspection by resident and region-based inspectors.

One Green finding was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

A Green self-revealing finding was identified for Entergy

=s failure to ensure the proper verification and calibration of vacuum trip switch VTS-1 during refueling outage (RFO) 16. Specifically, personnel did not ensure that the proper verification/calibration technique was employed to determine the as-found low condenser vacuum turbine trip setpoint. Additionally, when the technician identified that the as-found data was significantly outside of historical as-found values, he did not question the validity of the data nor did he obtain a peer check. The technician then calibrated the instrument using the incorrect as-found data which resulted in an incorrect low vacuum trip setpoint and a subsequent turbine trip and reactor scram on July 10, 2007.

This finding is more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective of limiting the likelihood of those events t hat upset plant stability during pow er operations. The finding is of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment would be unavailable. This finding has a cross-cutting aspect in the area of Human Performance,

Work Practices, because Entergy proceeded in the face of uncertainty or unexpected circumstances when the VTS-1 setpoint was found significantly outside of expected as-found values. H.4(a) (Section 4OA3)

B. Licensee-Identified Violations

None.

4

REPORT DETAILS

Summary of Plant Status

Pilgrim Nuclear Power Station (PNPS) operated at or near 100 percent power during the inspection period with the following exceptions: On October 30, 2007, Entergy reduced power to approximately 48 percent to perform a thermal backwash on the main condenser. Entergy resumed 100 percent power operation on October 31, 2007. On December 10, 2007, Entergy shut down and commenced a planned outage to repair leaking safety relief valve, RV-203-3B.

Entergy restored the plant to 100 percent power on December 13, 2007. The plant remained at or near 100 percent for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

a. Inspection Scope

(3 samples - 1 seasonal readiness, 2 impending adverse weather)

The inspectors performed a review of cold weather preparations during the onset of the cold weather season to evaluate the site's readiness for seasonal susceptibilities. The inspectors reviewed Entergy's preparations for cold weather and its impact on the protection of safety-related systems, structures and components (SSCs). The inspection focused on the intake structure, the station blackout diesel generator and the condensate storage and transfer system. The inspection was intended to ensure that Entergy's equipment, instrumentation, and supporting structures were configured in accordance with Entergy's procedures and that adequate controls were in place to ensure functionality of the systems in cold weather. The inspectors also conducted a site walkdown on November 1, 2007, to assess Entergy's readiness for the potential affects of hurricane Noel. The inspectors verified that all outside objects were properly anchored or tied down. In addition, the inspectors conducted a site walkdown on December 12, 2007, to evaluate site preparations for an approaching coastal storm with accompanying high winds.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

(4 samples)

The inspectors performed four partial system walkdowns during this inspection period.

The inspectors reviewed the documents listed in the Attachment to determine the correct 5system alignment. The inspectors conducted a partial walkdown of each system to determine if the critical portions of the selected systems were correctly aligned in accordance with these procedures and to identify any discrepancies that may have had an effect on operability. The walkdowns included selected switch and valve position checks, and verification of electrical power to critical components. Finally, the inspectors evaluated other elements, such as material condition, housekeeping, and component labeling. The following systems were reviewed based on their risk significance for the given plant configuration:

$ "B" Reactor Building Closed Cooling Water (RBCCW) system during degradation of the "A" RBCCW system;

$ "B" Residual Heat Removal (RHR) system during "A" RHR surveillance;

$ High Pressure Coolant Injection (HPCI) system while the Reactor Core Isolation Cooling (RCIC) system was out of service; and

$ RBCCW system "B" loop, upon restoration of "E" RBCCW pump following completion of 3.M.3-47.2, "'B' Train Functional Test of Individual Load Shed Component.

@

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown

a. Inspection Scope

(1 sample)

The inspectors completed a detailed review of the standby gas treatment (SBGT) system to verify the functional capability of the system. The inspectors conducted a walkdown of the system to verify that the critical components such as valves, switches, and breakers were aligned in accordance with procedures and to identify any discrepancies that could have an effect on operability.

The inspectors discussed system health with the system engineer and conducted a review of outstanding maintenance work orders to verify that the deficiencies did not significantly affect the SBGT system function. The inspectors also reviewed the condition report (CR) database to verify that equipment problems were being identified and appropriately resolved. In addition, the inspectors reviewed recent test results to ensure the air system leakage and charcoal filter efficiency met the requirements of Technical Specifications (TS) and procedures. Documents reviewed during the inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

61R05 Fire Protection (71111.05)

Fire Protection - Tours (71111.05Q)

a. Inspection Scope

(8 samples)

The inspectors performed walkdowns of eight fire protection areas during the inspection period. The inspectors reviewed Entergy's fire protection program to determine the required fire protection design features, fire area boundaries, and combustible loading requirements for the selected areas. The inspectors walked down these areas to assess Entergy's control of transient combustible material and ignition sources. In addition, the inspectors evaluated the material condition and operational status of fire detection and suppression capabilities, fire barriers, and any related compensatory measures. The inspectors then compared the existing conditions of the areas to the fire protection program requirements to ensure all program requirements were being met. Documents reviewed during the inspection are listed in the Attachment. The fire protection areas reviewed were:

$ Fire Zone 5.2, "B" Train Salt Service Water Pump Room;

$ Fire Zone 1.22, "B" Reactor Building Closed Cooling Water Pumps and Heat Exchanger Rooms;

$ Fire Zone 4.2, "B" Emergency Diesel Day Tank Room;

$ Fire Zone 4.4, "A" Emergency Diesel Day Tank Room;

$ Fire Area 1.9, Fire Zone 2.2, "A" Switchgear and Load Center Room;

$ Fire Area 1.9, Fire Zone 3.5, Vital Motor Generator Set Room;

$ Fire Zone 1.3, High Pressure Coolant Injection Pump/Turbine Room; and

$ Fire Zone 2.3, Battery Room A A.@

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

Internal Flooding Inspection

a. Inspection Scope

(1 sample)

The inspectors walked down selected areas of the plant including the cable spreading room, vital Motor Generator set, and HPCI pump room to assess the effectiveness of Entergy's internal flood control measures. The inspectors assessed the condition of watertight doors, floor sump systems, curbing, hatch and conduit seals, and floor drains.

The inspectors reviewed CR-PNP-2007-1020, "Review of NRC IN-2007-01, Recent Operating Experience Covering Hydrostatic Barriers," to determine whether Entergy was identifying internal flooding issues and taking appropriate corrective actions. The references used for this inspection are listed in the Attachment to this report.

b. Findings

No Findings of significance were identified.

1R11 Licensed Operator Requalification

.1 Resident Inspector Quarterly Review

a. Inspection Scope

(1 sample)

The inspectors observed licensed operator requalification training on November 6, 2007.

Specifically, the inspectors observed classroom Senior Reactor Operator (SRO) training on Emergency Planning, Emergency Action Level (EAL) Classification, and Protective Action Recommendation (PAR) procedures and processes. The inspectors assessed the training to determine if the training adequately prepared the SROs to determine EAL classification levels and to conduct PAR assessments. The inspectors reviewed the applicable training objectives to determine if they had been achieved. The inspectors verified that issues identified during the classroom session were entered into the corrective action program. Documents reviewed during the inspection are listed in the

.

b. Findings

No findings of significance were identified.

.2 Licensed Operator Requalification

a. Inspection Scope

(1 sample)

The following inspection activities were performed using NUREG 1021, Revision 9, "Operator Licensing Examination Standards for Power Reactors," Inspection Procedure 7111111, "Licensed Operator Requalification Program," Appendix A, "Checklist for Evaluating Facility Testing Material" and Appendix B, "Suggested Interview Topics."

The inspectors reviewed documentation of operating history since the last requalification program inspection. Documents reviewed included NRC inspection reports and licensee CRs that involved human performance issues. The purpose of the review was to ensure operational events that occurred during the last two years were not indicative of possible training deficiencies. The inspectors also discussed facility operating events with the resident staff.

The inspectors reviewed comprehensive written exams (these exams were administered in the fall, 2006), and the scenarios and job performance measures administered during the weeks of September 10 and 17, 2007, to ensure the quality of these exams met or exceeded the criteria established in the Examination Standards and 10 CFR 55.59, "Requalification." The inspectors observed the administration of the operating exams to two crews.

8Conformance with simulator requirements specified in 10 CFR 55.46, "Simulation Facilities" The inspectors observed simulator performance during the conduct of the examinations, and reviewed simulator discrepancy reports to determine whether facility staff was complying with the requirements of 10 CFR 55.46. The inspectors reviewed a sample of simulator tests including transients; normal and steady state; malfunctions; and core performance tests.

Conformance with operator license conditions The inspectors determined whether the operators were complying with the conditions of their license by reviewing the following:

$ five medical records (The records were complete; restrictions noted by the doctor were reflected on the individual's license; and physical exams were given within 24 months.);

$ eight proficiency watch-standing records and one reactivation record (Records indicated the licensed operators conformed with proficiency and reactivation watch-standing requirements of 10 CFR 55.53, Conditions of Licenses.); and

$ remediation training records for four licensed operators (These operators had failed either an annual operating test, a comprehensive written exam, or a requalification segment evaluation. The remediation records were acceptable.).

Licensee's feedback system The inspectors interviewed operator requalification instructors, training and operations management, and two licensed operators for feedback regarding the implementation of the licensed operator requalification program to ensure the requalification program was meeting their needs and responsive to their recommended changes.

On October 29, 2007, the inspectors conducted an in-office review of licensee requalification exam results. These results reflected the operators' performance on the annual operating tests; the comprehensive written exams were administered in the fall, 2006, and therefore those test results were not part of this in-office review. The inspector assessed whether pass rates were consistent with the guidance of NRC IMC 0609, Appendix I, "Operator Requalification Human Performance SDP." The inspectors verified that:

$ Crew failure rate on the dynamic simulator was less than 20 percent. (Failure rate was 0.0 percent)

$ Individual failure rate on the dynamic simulator test was less than or equal to 20 percent. (Failure rate was 0.0 percent)

$ Individual failure rate on the walkthrough test (job performance measures) was less than or equal to 20 percent. (Failure rate was 0.0 percent)9$ Individual failure rate on the comprehensive written exam was less than or equal to 20 percent. (As noted above, the comprehensive written exams were administered in the fall, 2006. Test results were previously documented in NRC IR 50-293/2006-005.)

$ More than 75 percent of the individuals passed all portions of the exam. (100% of the individuals passed all portions of the exam)

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

(2 samples)

The inspectors reviewed action plans for two SSC issues and reviewed the performance history of these SSCs to assess the effectiveness of Entergy

=s maintenance activities. The inspectors reviewed Entergy

=s CRs, corrective actions, and functional failure determinations made in accordance with Entergy procedures and the requirements of 10 CFR 50.65(a)(1) and (a)(2), ARequirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

@ In addition, the inspectors reviewed selected SSC classification, goals, corrective actions, performance criteria and monitoring plans to return the (a)(1) systems to (a)(2) status. Also, the inspectors selected a sample of system health reports for review to evaluate the results of system performance monitoring, material condition, and operations impact, to determine if actions taken were reasonable and appropriate. The references used for this inspection are listed in the to this report. The following issues were reviewed:

$ Turbine Controls Subsystem failure, failed maintenance rule performance criteria of one functional failure in two years (CR-PNP-2007-03673); and

$ A B@ Emergency Diesel Generator (EDG) exceeded maintenance rule performance criteria due to functional failures on October 25, 2006, and January 4, 2007 (CR-PNP-2007-0052).

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

(4 samples)

The inspectors evaluated online and shutdown risk management for emergent and planned activities. The inspectors reviewed maintenance risk evaluations, work schedules, and control room logs to determine if concurrent planned and emergent maintenance or surveillance activities adversely affect ed the plant risk already incurred with out-of-service components. The inspectors evaluated whether Entergy took the 10necessary steps to control work activities, minimize the probability of initiating events, and maintain the functional capability of mitigating systems. The inspectors assessed Entergy's risk management actions during plant walkdowns. Documents reviewed during the inspection are listed in the Attachment. The inspectors reviewed the conduct and adequacy of scheduled and emergent maintenance risk assessments for the following maintenance and testing activities:

$ Yellow risk condition during emergent unavailability of the "A" EDG due to an engine coolant leak in the turbo charger casing;

$ Vital Motor Generator Set maintenance;

$ Yellow Risk Condition during scheduled maintenance resulting in the unavailability of the HPCI system; and

$ Safety Relief Valve 3B pilot valve replacement outage shutdown risk assessment.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

(5 samples)

The inspectors reviewed five operability determi nations associated with degraded or non-conforming conditions to determine if the operability determination was justified and if the mitigating systems or those affecting barrier integrity remained available such that no unrecognized increase in risk had occurred. The inspectors also reviewed compensatory measures to determine if the compensatory measures were in place and were appropriately controlled. The inspectors reviewed licensee performance against related TS and Updated Final Safety Analysis Report (UFSAR) requirements. The inspectors reviewed the following degraded or non-conforming conditions:

$ CR-PNP-2007-03708, Mechanical Pressure Regulator (MPR) Setpoint Adjustment;

$ CR-PNP-2006-01802, Minimum Condensate Storage Tank Level to prevent Vortex formation at the HPCI/RCIC suction;

$ CR-PNP-2007-04172, EDG Fuel Oil Storage Volume;

$ CR-PNP-2007-04724, During the quarterly HPCI pump surveillance, the HPCI system did not achieve rated flow of 4250 gpm; and

$ CR-PNP-2007-04841, RHR pump P-203D revealed pump suction pressure drop outside acceptable range.

b. Findings

No findings of significance were identified.

111R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

(8 samples)

The inspectors reviewed eight samples of post-maintenance tests (PMT) during this inspection period. The inspectors reviewed these activities to determine whether the PMT adequately demonstrated that the safety-related function of the equipment was satisfied, given the scope of the work performed, and that operability of the system was restored. In addition, the inspectors evaluated the applicable test acceptance criteria to verify consistency with the associated design and licensing bases, as well as TS requirements.

The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution. Documents reviewed during the inspection are listed in the Attachment. The following maintenance activities and their post-maintenance tests were evaluated:

$ ACB-102 12-year Periodic Inspection & Maintenance, WO 51536960;

$ Salt Service Water Pump "D" Quarterly (TS/IST) Operability Test, WO 51535011;

$ Replace Bladder in T-223A with New Butyl Rubber Bladder, WO 51532443;

$ "A" EDG Turbocharger Replacement, WO 00129585;

$ HPCI MO-6, MO-35, MO-3 and MO-14 hydraulic lock modifications per MRs 51534480, 51534482, 51534483 and 51534484;

$ HPCI flow controller replacement per WO 0013195;

$ Repair/replace pilot valve on main steam Safety Relief Valve RV-203-3B; and

$ Source Range Monitor A B@ replacement per WO 51530724.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

(1 sample)

The inspectors reviewed shutdown and plant restart activities associated with a planned outage to replace the pilot on leaking Safety Relief Valve, RV-203-3B. The planned outage commenced on December 10, 2007, and was completed on December 12, 2007.

The inspectors reviewed Entergy

=s forced outage work schedule, risk evaluations, control room logs, and vessel cooldown and heatup rate data. The inspectors observed activities in the control room during the plant shutdown and startup. The inspectors conducted a walkdown of the primary containment to verify that there was no evidence of reactor coolant system leakage and that foreign material was being accounted for and controlled.

Documents reviewed during the inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

121R22 Surveillance Testing (71111.22)

a. Inspection Scope

(3 samples)

The inspectors reviewed three samples of surveillance activities to determine whether the

testing adequately demonstrated equipment operational readiness and the ability to perform the intended safety-related functions. The inspectors reviewed selected prerequisites and precautions to determine if they were met and if the tests were performed in accordance with the procedural steps. Additionally, the inspectors evaluated the applicable test acceptance criteria for consistency with associated design bases, licensing bases, and TS requirements. The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution.

Documents reviewed during the inspection are listed in the Attachment. The following surveillance tests were evaluated:

$ RCIC pump quarterly in-service test;

$ HPCI System Pump and Valve Quarterly and Biennial Comprehensive Operability; and $ Reactor Coolant System Leak Rate determination per TS 3/4.6.C, "Primary System Boundary Coolant Leakage."

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone:

Occupational Radiation Safety

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)

a. Inspection Scope

(9 samples)

During the period October 15-18, 2007, the inspector conducted the following activities to evaluate the operability and accuracy of radiation monito ring instrument ation, and the adequacy of the respiratory protection program relative to maintaining and issuing self-contained breathing apparatus (SCBA). Implementation of these programs was reviewed against the criteria contained in 10 CFR 20, "Standards for Protection Against Radiation;" applicable industry standards; and Pilgrim procedures.

The inspector reviewed the UFSAR to identify area, process, and emergency monitors that are installed at Pilgrim for the protection of workers. The inspectors reviewed the current calibration records for selected instrumentation, including the Turbine Building Radwaste Sump Area monitor (1815-8C), the Reactor Building 23' South East Access Area monitor (1815-2D), and the Reactor Building Outside Traversing In-Core Probe Room monitor (1815-2B).

The inspector selected hand-held radiation instruments, air monitors, contamination monitors, and electronic dosimeters currently in use in the plant, and reviewed the 13calibration records for this instrumentation. Included in this review were the calibration records for selected electronic dosimeters (DMC-2000), radiation survey instruments (RO-2, RO-2A, RO-20, Wide Range Telepole), contamination survey instruments (RM-14, MD-12, SAM-9), count room scalers (BC-4, SAC-4), and air samplers (H809V, Victoreen Lapel Sampler).

The inspector reviewed the maintenance records, safety interlock checks, and current calibration source activity/dose rate determinations for the Shepard Model 78, Shepard Model 423, and Model 773 instrument calibrators.

The inspector evaluated the licensee's program for assuring quality in the radiation monitoring instrumentation and respiratory protection programs by reviewing 16 CRs related to radiation instrumentation, SCBA's, and the monitoring of plant radiation levels to determine if problems were identified in a timely manner and appropriate corrective actions were taken to resolve the related issues.

There were no incidents of personnel internal exposure resulting in a Committed Effective Dose Equivalent > 50 mrem that would require an in-depth evaluation of whole body counting instrumentation and bioassay techniques.

The inspector reviewed actions for radiation worker and radiation protection technician errors to determine whether the corrective actions were adequate to prevent recurrence.

The inspector verified calibration due dates and observed a technician performing source checks on a variety of instruments including portable radiation survey instruments (RO-2, Wide Range Telepole), contamination survey instruments (RM-14s, SAM 9), count room scalers (BC-4), and personal contamination monitors (PPM-1, PM-7).

The inspector reviewed surveillance records for ten SCBAs staged for use in the control room, Radiological Controlled Area access location, and the fire brigade equipment staging area in the fire service pump building. The inspector observed a technician perform an inspection of six of the ten units staged for use. The inspector observed a technician fill two SCBA air bottles from the air compressor unit. The sample results for breathing air, used to refill the SCBA tanks, were reviewed to confirm that air quality met CGA-G-7.1-2004 Grade D standards.

The inspector evaluated the adequacy of the respiratory protection program regarding the issuance of SCBAs to workers. Training and qualification records for licensed operators, radiation protection technicians, and fire brigade members required to wear SCBA's, in the event of an emergency, were reviewed.

b. Findings

No findings of significance were identified.

142PS3 Radiological Environmental Monitoring Program and Radioactive Material Control Program (71122.03)

a. Inspection Scope

(1 sample)

During the period October 15-18, 2007, the inspector conducted the following activity to determine whether the licensee's surveys and controls are adequate to prevent the inadvertent release of licensed materials into the public domain. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, "Standards for Protection Against Radiation;" TS; and Entergy procedures. This inspection activity represents completion of one sample relative to this inspection area.

The inspector observed the radioactive material survey and release locations. The methods used for control, survey, and release were inspected and included observations of the performance of personnel surveying and releasing material for unrestricted use and verifying that the work is performed in accordance with plant procedures.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

[OA]

4OA1 Performance Indicator (PI)

.1 Mitigating System Cornerstone

a. Inspection Scope

(2 samples)

The inspectors sampled data for the Mitigating System Performance Index PIs for the EDGs and cooling water systems (Salt Service Water and RBCCW) for the 4th quarter 2006 and 1st, 2nd and 3rd quarter 2007 to assess the completeness and accuracy of the reported information. The inspectors reviewed operator logs, CRs, maintenance rule documents, maintenance records, Licensee Event Reports (LERs), system health reports, and plant process computer information. The acceptance criteria used for the review were Nuclear Energy Institute (NEI) 99-02, Revision 5, "Regulatory Assessment Performance Indicator Guidelines."

b. Findings

No findings of significance were identified.

.2 Physical Protection Cornerstone

a. Inspection Scope

(3 samples)

The inspectors performed a review of PI data submitted by the licensee for the Physical Protection Cornerstone. The review was conducted of the licensee

=s programs for gathering, processing, evaluating, and submitting data for the Fitness-for-Duty, Personnel 15Screening, and Protected Area Security Equipment PIs. The inspectors determined whether the PIs had been properly reported as specified in NEI 99-02. The review included the licensee

=s tracking and trending reports, personnel interviews, and security event reports for the PI data collected since the last security baseline inspection. The inspector noted from the licensee

=s submittal that there were no reported failures to properly implement the requirements of 10 CFR 73, "Physical Protection of Plants and Materials," and 10 CFR 26, "Fitness for Duty Programs," during the reporting period. This inspection activity represents the completion of three samples relative to this inspection area; completing the annual inspection requirement.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered into the Corrective Action Program (CAP)

a. Inspection Scope

The inspectors performed a screening of each item entered into the licensee's CAP. This review was accomplished by reviewing printouts of each CR, attending daily screening meetings and/or accessing the licensee's database. The purpose of this review was to identify conditions such as repetitive equipment failures or human performance issues that might warrant additional follow-up.

b. Findings

No findings of significance were identified.

.2 Semi-Annual Review to Identify Trends

a. Inspection Scope

(1 sample)

As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"

the inspectors performed a review of Entergy

=s CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment and corrective maintenance issues but also considered the results of daily inspector CAP item screening discussed in Section 4OA2.1. The review also included issues documented in CAP trend reports and site CAP performance indicator data. The inspectors review considered the six month period of June through December, 2007, although the inspectors also evaluated the trend review results discussed in NRC IR 05000293/2007003, which reviewed CRs from October 2006 through May 2007. Documents reviewed during the inspection are listed in the

.

16b. Assessment and Observations No findings of significance were identified. The inspectors noted a number of plant equipment configuration control issues discussed in the third quarter 2007 Pilgrim Station Quarterly Trend Report, including:

$ CR-PNP-2007-00303, PS-CKVS-B (Crankcase Pressure Switch A B@ Diesel) not valved in correctly;

$ CR-PNP-2007-01446, RCIC check valve 1301-CK-50 initial position found open instead of closed;

$ CR-PNP-2007-02383, Breaker B1446 (EDG A B@ Diesel Oil Transfer Pump) found A OFF,@ normal position is A ON;@ $ CR-PNP-2007-02468, Isolation valve found closed on Reactor Pressure Transmitter;

$ CR-PNP-2007-02476, Spare breaker found closed when it was expected to be open; and

$ CR-PNP-2007-02651, EDG failed to start (likely due to fuel rack and governor left in full fuel position).

The report concluded that the number of issues Adoes not exhibit an adverse or emerging trend,@ but that Operations Management considers the number of Amispositionings

@ to be at an unacceptable level. The inspectors also considered the number of issues discussed in the report to be at an unacceptable level, however, the inspectors also concluded that these issues represent a low level trend in the area of configuration control. The inspectors have discussed this trend with licensee management and will continue to monitor configuration control issues at Pilgrim during this assessment period.

.3 Annual Sample: Review of Outage CRs

a. Inspection Scope

(1 sample)

The inspectors reviewed a sample of CRs from Pilgrim's 2007 refueling outage to determine whether CRs initiated during the outage were processed and closed in accordance with Pilgrim procedures. The inspectors reviewed two Apparent Cause Evaluations conducted by Pilgrim. The inspectors evaluated whether corrective actions taken by Pilgrim addressed each CR as well as the overall process. Documents reviewed are listed in the Attachment.

b. Assessment and Observations No findings of significance were identified. The inspectors determined that there were many instances where the condition review group (CRG) closed a lower level (Category D) CR to "supervisory oversight." Managers would perform follow-up and close the CR with a general statement such as "Corrective actions for the CR were reviewed by the responsible manager. Upon the manager's recommendation, this CR is being closed."

This practice resulted in a condition where corrective actions for a particular issue could not be tracked or demonstrated. Pilgrim has since discontinued this practice as an acceptable closure strategy for Category D CRs.

17.4 Annual Sample: Review of Motor Operated Valve (MOV) Hydraulic Lock

a. Inspection Scope

(1 sample)

The inspectors selected CR-PNP-2006-04328 for detailed review. The CR was written to determine the cause of a safety-related MOV failure in the RHR system during routine surveillance testing. The inspectors reviewed the licensee's root cause analysis, corrective actions, and the prioritization of the corrective actions.

b. Assessment and Observations No findings of significance were identified. The inspectors determined that the licensee performed a thorough root cause analysis and took timely corrective actions to prevent recurrence. The root cause was determined to be hydraulic locking of the MOV actuator due to grease found inside of the spring package. The grease prevented the spring package from compressing which in turn prevented the thermal overloads from tripping.

The tripping of the thermal overloads stops the motor and provides the indication that the valve is closed.

The root cause analysis determined that newer MOVs in the plant were not susceptible to hydraulic lock because the valves have an internal grease relief path from the spring package to the actuator housing. However, most MOVs at Pilgrim did not have the internal grease relief path. Immediate corrective actions included looking inside the spring package of all safety-related MOVs for grease. Long term corrective actions for this issue included a design modification to provide an external grease relief path from the spring package back to the actuator housing. All of the high priority valves have been modified. The last low priority valve to receive this modification is scheduled to be performed in the next refueling outage. The inspectors determined that the prioritization of the corrective actions was appropriate.

.5 Annual Sample: Follow-up Review of Component Design Bases Inspection (CDBI)

Finding Regardi ng the Inadequate Operability Determination for the HPCI Turbine Trip Solenoid Failure

a. Inspection Scope

(1 sample)

The inspectors reviewed the corrective actions for a finding identified during the CDBI and documented in inspection report number 05000293/2006006. The finding was associated with Entergy

=s failure to declare the HPCI system inoperable due to a HPCI turbine trip solenoid failure. The inspectors reviewed CR-PNP-2006-01460 to determine whether the corrective actions were appropriate and completed. As part of this review, the inspectors examined various safety system operating procedure changes to assess their adequacy.

The documents reviewed are listed in the Attachment to this report.

b. Assessment and Observations

No findings of significance were identified. Entergy

=s initial failure to declare HPCI inoperable was due to licensing and operations department management focusing on the ability of the HPCI system to perform its accident analysis function versus a discussion of 18the HPCI system TS requirements. The fo cus did not address the ability of the HPCI system to automatically trip on high water level in the reactor vessel, as described in TS 3.2.B, "Protective Instrumentation Core and Containment Cooling Systems - Initiation and Control." As a result, the HPCI system should have been considered inoperable regardless of the ability of the system to perform its accident analysis function.

The inspectors determined that the licensee

=s corrective actions were appropriate. Entergy determined the failure to declare HPCI inoperable was due to a lack of independence of the operations department and licensi ng departments in reviewing operability determinations. The inspectors noted that Entergy immediately implemented operations department training regarding independent review of emerging TS issues. Also, Entergy revised safety system operating procedures to include a section on TS instrumentation requirements.

4OA3 Event Follow-up

.1 Infrequently Performed Evolution: MG Set Power Transfer

a. Inspection Scope

(1 sample)

On October 3, 2007, Pilgrim operators performed a planned manual transfer of vital alternating current (AC) power from its normal power source, the vital MG set, to its alternate power source, bus B15, with the plant at power. This infrequently performed evolution was conducted to remove the vital MG set from service for repairs. The evolution posed several challenges to Pilgrim operators because the transfer of the vital AC power from its normal to its alternate source would cause a momentary interruption in vital AC power. Similar evolutions in the past had resulted in complications such as the receipt of reactor building isolation signals, feed regulating valve position lock ups, and recirculation pump scoop tube position lock ups. Entergy developed a new procedure for this evolution, Procedure 2.2.16, Attachment 8, "A Manual Transfer of Y2 to Motor Control Center (MCC)

B15 with the Units On-line." The procedure established several compensatory measures to mitigate the effects of a component malfunction or unexpected response. For instance, operators were briefed on Procedure 2.4.49, Section 4.4, "A Manual Lockup of Feed Regulating Valve(s) from the Condenser Bay," and were stationed outside the condenser bay to take manual control of the valves if needed. Additionally, operators inserted a reactor building isolation signal before the vital power transfer, to prevent the signal from coming in during the transfer. The inspectors reviewed the procedure and observed the evolution from the control room to assess operator actions, command and control, and the adequacy of communications within the control room and between the control room and the field.

b. Findings

No findings of significance were identified.

19.2 LER Review and Closeout (1 sample)

(Closed) LER 05000293/2007-005-00, Reactor Scram Resulting from Low Vacuum Turbine Trip

a. Inspection Scope

The inspectors reviewed Entergy

=s actions associated with LER 50-293/2007-05-00, which discussed the July 10, 2007, low vacuum turbine trip and automatic reactor scram event. The inspectors reviewed the licensee

=s LER and associated root cause evaluation. Additionally, the inspectors verified that follow-up actions, taken or planned, were appropriate to address the event. This LER is closed.

b. Findings

Introduction:

A Green self-revealing finding was identified for Entergy

=s failure to ensure the proper verification and calibration of vacuum trip switch VTS-1 during refueling outage (RFO) 16. Specifically, personnel did not ensure that the proper verification/calibration technique was employed to determine the as-found low condenser vacuum turbine trip setpoint. Additionally, when the technician identified that the as-found data was significantly outside of historical as-found values, he did not question the validity of the data nor did he obtain a peer check. The technician then calibrated the instrument using the incorrect as-found data which resulted in an incorrect low vacuum trip setpoint and a subsequent turbine trip and reactor scram on July 10, 2007.

Description:

On July 10, 2007, an unplanned automatic reactor scram occurred while performing condenser thermal backwashes at approximately 48 percent power. The reactor protection system (RPS) scram signal was initiated by the trip of the main turbine on low condenser vacuum. Pilgrim operators stabilized t he plant in a shutdown condition and made a four-hour notification to the NRC. Post scram review of the as-found setpoint for vacuum trip switch, VTS-1, revealed that the trip setpoint was set to actuate at 24.35

@ Hg rather than the expected 21.95

@ B 22.45@ Hg. Entergy recalibrated the vacuum switch and restored the plant to 100 percent power on July 16, 2007.

Entergy conducted a root cause evaluation of the unplanned scram and summarized their results in LER 2007-005-00, "Reactor Scram Resulting from Low Vacuum Turbine Trip."

Entergy determined that the root cause of the event was that the technician who had calibrated the VTS-1 switch during RFO 16 had not properly implemented human performance tools (e.g., training) for this particular type of large volume instrument to ensure a proper calibration. Specifically, since the bellows for VTS-1 are very large, the vacuum must be decreased slowly during the calibration in order for an accurate setpoint to be obtained. While obtaining the as-found setpoint, the technician did not decrease the vacuum slowly which resulted in faulty as-found results. Additionally, when the as-found data suggested that the vacuum switch was considerably outside of historical results, the technician did not question the validity of the data nor did he obtain a peer check. The technician then made adjustments to the instrument using the incorrect as-found data.

20Entergy=s root cause report also discussed several weaknesses with Procedure 8.F.51, ATurbine Generator and Auxiliary Instruments Calibration.

@ Specifically, the root cause report noted that Aadditional details in the procedure would provide an additional barrier to ensure the proper calibration technique is achieved.

@ However, the inspectors noted that Entergy had not identified these procedural weaknesses as a contributing cause to this event. The inspectors concluded that the lack of procedural specificity and guidance contributed to the improper calibration of VTS-1. Entergy

=s corrective actions for this aspect included adding steps to the procedure to decrease the vacuum at a slower rate, to include detailed guidance on the adjustments of the trip and span of the vacuum trip assembly, and to require supervisory review of as-found data and testing techniques prior to performing adjustments.

Analysis:

The performance deficiency associated with this finding is that Entergy did not ensure the proper verification and calibration of vacuum trip switch VTS-1 during RFO 16.

The improper setpoint resulted in a low vacuum turbine trip and consequent automatic reactor scram on July 10, 2007.

This finding is more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and affects the

cornerstone objective of limiting the likelihood of those events that upset plant stability during power operations. The inspectors conducted a Phase 1 screening in accordance with IMC 0609, "Significance Determination Process," Appendix A, "Reactor Inspection Findings for At-Power Situations." The finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment would be unavailable. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because Entergy proceeded in the face of uncertainty or unexpected circumstances by continuing with the calibration procedure even though the vacuum trip switch setpoint was found significantly outside of historical as-found values. H.4(a)

Enforcement:

Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement in that the vacuum trip switch is not a safety-related component. Entergy has entered this issue into their corrective action program as CR-PNP-2007-3231. Corrective actions included recalibrating VTS-1 before the plant restart, providing remedial training for the technician who had conducted the improper calibration, and adding vacuum switch fundamentals as a continuing training topic for the instrumentation and controls (I&C) technicians. Additional corrective actions planned by Entergy include revising Procedure 8.F.51 to include more detailed guidance and to require a supervisory review of as-found data prior to performing adjustments; conducting just-in-time training prior to the RFO 17 vacuum trip switch setpoint verification and calibration; and identifying and revising other I&C procedures involving critical calibrations. Because this violation does not involve a violation of regulatory requirements and has a very low safety significance, it is identified as FIN 05000293/2007005-01, Improper Calibration of Vacuum Trip Switch Results in an Automatic Reactor Scram.

4OA6 Meetings, Including Exit

On October 18, 2007, an Occupational Radiation and Public Radiation Safety exit meeting was conducted. The preliminary inspection results were presented to Robert Smith, 21General Manager Pilgrim Operations, and other members of the Pilgrim staff. The licensee did not identify any material as proprietary during this inspection.

On October 18, 2007, the Security inspection results were presented to members of licensee management.

On January 9, 2008, the resident inspectors conducted an exit meeting and presented the preliminary inspection results to Mr. Kevin Bronson, Site Vice President, and other members of the Pilgrim staff. The inspectors confirmed that no proprietary information was provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

S. Bethay Nuclear Safety Assurance Director

K. Bronson Site Vice President, Pilgrim

H. Bouska Supervisor, Operations Training

D. Burke Security Manager

L. Foreaker Supervisor, Radiation Instrumentation

J. Henderson Manager, Radiation Protection

M. Gakka Licensing

T. Kelly Technician, Radiation Protection

R. Larson Technician, Radiation Protection

W. Lobo Licensing Engineer

J. Lynch Licensing Manager

F. Marcussen Protective Services Department Manager

C. McMorrow Senior Operations Instructor

D. Noyes Operations Director

M. Santiago Superintendent, Nuclear Training

L. Seehaus Technician, Radiation Protection

R. Smith Plant Operations General Manager

D. Towmey Lead Technician, Radiation Protection

LIST OF ITEMS

Opened and Closed

05000293/2007005-01 FIN Improper Calibration of Vacuum Trip Switch Results in an Automatic Reactor Scram

Closed

05000293/LER-2007-005-00 LER Reactor Scram Resulting from Low Vacuum Turbine Trip

LIST OF DOCUMENTS REVIEWED

Section 1R01

UFSAR Table 10.9-1, Design Temperatures

NRC

IN 96-036, Degradation of Cooling Water Systems Due to Icing

NRC

IN 98-002, Nuclear Power Plant Cold Weather Problems and Protective Measures

Procedure

8.C.40, Seasonal Weather Surve illance, Attachment 1, Cold Weather Preparations, Revision 19 Procedure 2.2.35, Condensate Storage and Transfer System, Revision 40

Section 1R04

Drawing M215 Sheet 2, Revision 48, P&ID Cooling Water System Reactor Building Drawing M215 Sheet 5, Revision E8, Composite P&ID Cooling Water System Reactor Building

Procedure

2.2.30, Revision 65, RBCCW System

CR-PNP-2007-04299

Procedure

2.2.19, Residual Heat Removal System, Revision 95

M241, P21D, Residual Heat Removal System, Revision 47

PNPS Procedure 2.2.21, Revision 72, High Pressure Coolant Injection System

Procedure

7.1.44, "Sampling of Charcoal Cells in SBGT and Control Room Environmental Filters' Systems for Methyl Iodide Testing", completed on 11/28/06 for "B" SBGT

LO-NOE-2007-00092

PNPS Procedure 2.2.50, SBGT

PNPS Drawing M294, Heating Ventilation and Air Conditioning SBGT System Control Diagram, Revision 16

WO 05106023, Leak Rate Test of Air Supply for SBGT System Dampers, 10/2/07

PNPS Procedure 8.M.2-7.1.19, Revision 4, Attachment 4, "Allowable Daily Leakage Rate"

PNPS Final Safety Analysis Report, Revision 10, Chapter 5.3.3.4, SBGT System

PNPS Final Safety Analysis Report, Revision 10, Chapter 7.18, Reactor Building Isolation and Control System

CR-PNP-2007-03013

Pilgrim TS 3.7.B, SBGT System and Control Room High Efficiency Air Filtration System

Procedure

2.2.30, RBCCW System, Revision 65

Procedure

3.M.3-47.2, A B@ Train Functional Test of Individual Load Shed Components,

Revision 18

Section 1R05

Pre Fire Plan, Screenhouse Building EL. 23' Pre Fire Plan, Reactor Building Quads, EL. 17'6"

89XM-1-ER-Q, Updated Fire Hazards Analysis, Revision E5

Procedure

5.5.2, Special Fire Procedure, Revisions 29 and 37

PNPS Procedure 8.B.17.2, Inspection of Fire Damper Assemblies, Attachment 1, Revision 9, completed 4/3/07 PNPS Procedure 8.B.17.2, Inspection of Fire Damper Assemblies, Attachment 11, Revision 9, completed 4/4/07

Section 1R06

PNPS-PSA, Revision 1, PNPS Probabilistic Safety Assessment IPE Update NRC

IN 2007-01, Recent Operating Experience Concerning Hydrostatic Barriers

Procedure

3.M.4-96, Floor Plug and Vault Hatch Seals

CR-PNP-2007-01020,

CR-PNP-2006-03750,

CR-PNP-04223,

CR-PNP-00312,

CR-PNP-01123,

CR-PNP-02708,

CR-PNP-03457

Section 1R11

Lesson Plan O-RO-07-02-01, Revision 4, Emergency Classification and Notification

NRC

RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events

EP-IP-100, Revision 26, Emergency Classification and Notification

A-3EP-IP-300, Revision 6, Offsite Radiological Dose Assessment

EP-IP-400, Revision 11, PARs Lesson Plan O-RO-07-03-03, Revision 0, PARs,

EP-IP-400

CR-PNP-2007-4587, Control Room does not have the same weather assessment capability (160' Met Tower) for EAL assessment as the EOE

CR-PNP-2007-4591,

EP-IP-400 states that core temperature >2400F is indication of substantial core damage, this temperature is not able to be obtained Section 1R12

EN-DC-203 R0, MR Program

EN-DC-204 R0, MR Scope and Basis

EN-DC-205 R0, MR Monitoring

EN-DC-206 R0, MR (a)(1) Process

EN-LI-102 R9, CA Process

EN-LI-121 R6, Entergy Trending Process

CR-PNP-2007-00552

A B@ EDG exceeded MR reliability performance criteria

CR-PNP-2007-03849 CA1 Functional Failure Determination Form (9/3/07)

CR-PNP-2007-03673 Turbine Controls System (a)(1) Action Plan

CR-PNP-2007-00552

A B@ EDG (a)(1) Action Plan Health Report, System 02, Reactor Recirculation 3

rd Qtr 2007 Health Report, System 29, Salt Service Water 3

rd Qtr 2007 Health Report, System 01, Main Steam, 3

rd Qtr 2007 10/09/2007, MR Expert Panel Meeting Minutes Section 1R13

Risk Management Actions

CR-PNP-2007-04579, Small leak observed at the base of the "A" EDG turbo charger gas inlet casing Procedure 2.2.16, Revision 50, Attachment 8, Manual Transfer of Y2 to MCC B15 with the unit on-line TS 3.5.C.2, HPCI System Equipment out of service (EOOS) quantitative risk assessment tool

Procedure

3.M.1-45, Outage Shutdown Risk Assessment, Revision 6

Risk Assessment Review Checklist for 12/10 08:00 to 12/12 18:00

EOOS Scheduler

=s Evaluation for PNPS for 12/10 0:00 to 12/13 12:00 Risk Assessment Review Checklist for 12/10 08:00 to 12/12 18:00, Revision A

Section 1R15

CR-PNP-2007-03708, Adjustments of the MPR Setpoint have been required.

ODMI Action Plan for MPR Setpoint Adjustments Apparent Cause Evaluation for MPR Setpoint Drifting

CR-PNP-2006-01802

CR-PNP-2007-04172

Operability Determination for CR-

PNP-2007-04172 Procedure 8.9.1, Revision107, Attachment 3, EDGs On-Site Fuel Oil Quantity

TS 3.9.A, Revision 212, Auxiliary Elec trical Equipment TS 3.5.C, HPCI System

TS 3.12, Fire Protection, Alternate Shutdown Panels

CR-PNP-2007-04724, HPCI did not achieve rated flow during operability testing 50.72 Notification for loss of HPCI Safety Function Entergy procedure

ENN-OP-104, AOperability Determinations

@

CR-PNP-2007-04841, Initial operability review for pump P203D

pump suction pressure drop value not acceptable

CR-PNP-2007-04871, LPCI system loop

A B@ pump and valve quarterly operability Procedure 8.5.2.2.2, LPCI system loop

A B@ Operability-Pump Quarterly and Biennial (Comprehensive) Flow Rate Tests and Valve Tests

51535468 01, Work Order, LPCI system loop

A B@ PP V1v Quarterly Operability P203D Test Date Sheet, RHR Inservice pump test data sheets for 11/26 and 12/3/2007

Section 1R19

Procedure

8.5.3.2.1, Revision 19, Attachment 1D, Quarterly and Biennial (Tech Spec/IST) Test Procedure for SSW pump D (P-208-D)

CR-PNP- 2007-04274,

CR 2007-04251,

CR 2007-04264

Apparent Cause Evaluation for

CR 2007-4274

WO 51532443, Replace Bladder in T-223A with New Butyl Rubber Bladder, 10/11/07

WO 00129585, "A" EDG, Leakage Observed from Base of Turbocharger,

CR 2007-04172

CR-PNP-2007-04724

M1J18-11, Elementary Diagram High Pressure Coolant Injection System

4533K40-800, page 43/44, Figure 24: Schematic Diagram of Type 540-01 and 540-51 controller (for HPCI flow controller)

MR 51534480 - Install Hydraulic Lock Modification for HPCI

MO-6

MR 51534482 - Install Hydraulic Lock Modification for HPCI

MO-35

MR 51534483 - Install Hydraulic Lock Modification for HPCI

MO-3

MR 51534484 - Install Hydraulic Lock Modification for HPCI

MO-14

Procedure

1.3.34, Operations Administrative Policies and Processes, Revision 113

Procedure

2.2.21.5, HPCI Injection and Pressure Control, Revision 13

Procedure

8.5.4.1, HPCI System Pump and Valve Quarterly and Biennial Comprehensive Operability, Revision 102

Procedure

8.5.4.4, HPCI Valve (Quarterly) Operability Test, Revision 48 Procedure 8.E.23, HPCI System Instrument Calibration, Revision 65

Procedure

8.M.2-2.5.7, Instrument Functional/Calibration Test For HPCI Suppression Chamber Water Level, Revision 49

WO 00131058, HPCI Injection Flow Controller

50.72 Event Report to USNRC: High Pressure Coolant Injection Inoperable, dated November 20, 2007 Control Room (day) Shift Narrative Logs, dated November 19, 2007

LER 2000-002-00, AHigh Pressure Coolant Injection System Inoperable due to Power Inverter Failure@ TS 3.12, Fire Protection, Alternate Shutdown Panels

WO 00125819, Source Range Monitor (SRM) Discriminator (SRM B)

WO 00133189, SRM B Neutron Flux Response Functional Test

CR-PNP-2007-04937, Air leakage identified at connection between the solenoid valve and the manifold Procedure 3.M.4-6, Removal, installation, Test, Disassembly, Inspection, and Reassembly of Main Steam Relief Valves 3379-270-3 E5, Main Steam SRV Sheets 1, 2, 3 and 4

A-53379-271-1 E1, Main Steam SRV Parts List Sheets 1, 2, 3, and 4

WO 51535014, WO

RV-203-3B Tailpipe temperature has trended up-pilot valve change out

WO 00133198, WO Automatic Depressurization System subsystem manual opening of relief valves Procedure 2.1.19, Suppression chamber temperatures

Procedure

8.5.6.2, Special test for ADS system manual opening of relief valves

Section 1R20

PNP On-Line Master Schedule, dated 11/30/07, 12/10/07, and 12/11/07

Procedure

2.1.5, Controlled Shutdown from Power, Revision 103

Procedure

2.2.19.1, Residual Heat Removal System - Shutdown Cooling Mode of Operation, Revision 24 Procedure 2.1.1, Startup from Shutdown, Revision 162

Procedure, 2.1.7, Vessel Heatup and Cooldown, Revision 52, completed 12/12/2007

Section 1R22

Procedure

8.5.5.1, Revision 56, RCIC Pump Quarterly and Biennial Operability Flow rate and Valve Test at approximately 1000 psig

WO 51534877, RCIC Pump Operability and Flow Rate Te st at 1000 psig, 10/10/07

CR-PNP-2007-04640;

CR-PNP-2007-04816;

CR-PNP-2007-04835

Procedure

6.1-220, Radiological Controls for High Risk Evolutions, Revision 2

Procedure

8.I.1.1, Inservice Pump and Valve Te sting Program, Revision 21 Procedure 8.5.4.1, HPCI System Pump and Valve Quarterly and Biennial Comprehensive Operability, Revision 102

EN-RP-131, Attachment 9.2, Revision 3, Air Sampling results from November 11, 2007

EN-RP-131, Attachment 9.2, Revision 3, Air Sampling results from November 19, 2007

Control Room (day) Shift Narrative Logs, dated 11/20/2007

Technical Specification 3.5.C, High Pressure Coolant Injection

UFSAR Section 6.5.2.3, High Pressure Coolant Injection System

USNRC Letter to Entergy: PNPS - Entergy Relief Request

PR-03 High Pressure Coolant Injection Pump, dated August 29, 2005

Procedure

2.1.15, Daily Surveillance Log, Revision

Procedure

8.M.2-5, Drywell Drain Sump Integrator, Revision 9, Attachment 1, completed 10/18/07

Procedure

8.M.2-5, Drywell Drain Sump Integrator, Revision 9, Attachment 2, completed 10/6/05

Drawing C-75, Reactor Building Foundations Drywell Concrete @ El. 9'-2, Revision 4

ER#

06110910, Attachment 9.1

Control Room Shift Narrative Logs, dated 12/5/2007 through 12/7/2007

Sections 2OS1/2OS2/20S3

6.5-003, Revision 8, Radiation Protection Instrumentation Calibration Frequency

6.5-160, Revision 31, Calibration of the Area Radiation Monitoring System

6.5-170, Revision 21, Calibration of Ventilation System Radiation Monitors Using ARM Type Sensor/Converters 6.5-307, Revision 16, Calibration of the Eberline

RO-2/RO2A or

RO-20 Ion Chamber

6.5-311, Revision 10, Calibration of the Eberline Model

RO-7 Radiation Monitor

6.5-341, Revision 11, Calibration of the

MDC 2000S Electronic Dosimeter

6.7.1-106, Revision 14, Inspection and Testing of Respiratory Protection Equipment

6.7.1-201, Revision 8, Operation of the SCBA Air Compressor

A-6EN-RP-121, Revision 1, Radioactive Material Control

EN-RP-301, Revision 0, Radiation Protection Instrument Control

EN-RP-303, Revision 0, Source Checking of Radiation Protection Instrumentation

EN-RP-502, Revision 1, Inspection and Maintenance of Respiratory Protection Equipment Calibration Records:

Electronic Dosimeter Calibration (Serial Nos.

176631,

219267,

178032,

177025, 170628)

E-520 (Serial No. 722)

SAC-4 (Serial No. 1402)

BC-4 (Serial No. 484)

Victoreen Lapel Sampler (Serial No. c1138)

H809V (Serial No. 6168)

PM-7(Serial No. 600, 392)

Wide Range Telepole (Serial No. 6603-027)

RO-2 (Serial No. 3410)

RO-2A (Serial No. 3295)

RO-20 (Serial No. 325, 285)

RO-7 (Serial No. 1030)

RM-14 (Serial No. 8565)

SAM-9 (Serial No. 308)

MD-12 (Serial No. 135005)

CR-PNP-2007-00078, 00426, 01012, 01077

CR-PNP-2006-00844, 01290, 01792

CR-PNP-2007-00341, 01372, 03317

CR-PNP-2006-00620, 00843, 01432, 03085, 03922, 03935

SCBA Numbers :1, 2, 3, 4, 5, 10, 11, 12, 13, 14

Miscellaneous

Records & Reports:

Mask Qualification List Root Cause Analysis Report for

CR-PNP-07-3880

Instructional Module C-FB-02-02-01, Revision 7 Self-Contained Breathing Apparatus

Section 4OA2

Limitorque Maintenance Update 90-1

Limitorque Maintenance Update 88-2

ER 07101434, Revision 0, Installation of External Grease Relief Bypass on Limitorque Actuators

ER 07112191, Revision 0, Revision to

VM-0390 to Provide Additional Instructions for Installation of MOV External Grease Relief Modifications

DRN 07-01007, Limitorque Valve Controls Third quarter 2007 Pilgrim Station Quarterly Trend Report

NRC

IR 2007-003

CR-PNP-2007-03925, Potential Adverse Trend in Station Mispositioning errors

CR-PNP-2007-00303,

PS-CKVS-B (crankcase pressure switch

A B@ diesel) not valved in correctly

CR-PNP-2007-01446, RCIC check valve 1301-CK-50 initial position found open instead of closed

CR-PNP-2007-2383, Breaker B1446 (EDG

A B@ Diesel Oil Transfer Pump) found

A OFF@, normal position is

A ON@

CR-PNP-2007-02468, Isolation valve found closed on Reactor Pressure Transmitter

CR-PNP-2007-02476, Spare breaker found closed when it was expected to be open

CR-PNP-2007-02651, EDG failed to start (likely due to fuel rack and governor left in full fuel position)

A-7Procedure 2.2.21, High Pressure Coolant Injection System, Revision 72 Procedure 2.2.19, Residual Heat Removal System, Revision 95

Procedure

2.2.3, Automatic Depressurization System, Revision 23

Procedure

2.2.8, Emergency Diesel Generator, Revision 90

LIST OF ACRONYMS

AC alternating current

ADAMS Agencywide Documents Access and Management System

CAP corrective action program

CDBI component design bases inspection

CFR Code of Federal Regulations

CR condition report

CRG Condition Review Group

DRP Division of Reactor Projects

DRS Division of Reactor Safety

EAL emergency action level

EDG emergency diesel generator

gpm gallon per minute

Hg mercury

HPCI high pressure coolant injection

I&C instrumentation and controls

IMC Inspection Manual Chapter

IR Inspection Report

LER Licensee Event Report

MCC motor control center

MG motor generator

MO motor-operated

MOV motor-operated valve

MPR mechanical pressure regulator

mrem millirem

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

PAR Protective Action Recommendation

PARS Publicly Available Records

PI Performance Indicator

PMT post-maintenance test

PNPS Pilgrim Nuclear Power Station

RBCCW reactor building closed cooling water

RCIC reactor core isolation cooling

RCS reactor coolant system

RFO refueling outage

RHR residual heat removal

RV relief valve

RPS reactor protection system

SBGT stand by gas treatment

SCBA self-contained breathing apparatus

A-8SDP Significance Determination Process

SRM source range monitor

SRO senior reactor operator

SRV safety relief valve

SSC system, structure, or component

SSW salt service water

TS Technical Specifications

UFSAR Updated Final Safety Analysis Report

URI unresolved item

VTS vacuum trip switch

WO work order