IR 05000293/2012002

IR 05000293-12-002; Entergy Nuclear Operations, Inc; 01/01/2012 - 03/31/2012; Pilgrim Nuclear Power Station, Maintenance Effectiveness and NRC Integrated Inspection Report

ML12116A289
Person / Time
Site:	Pilgrim
Issue date:	04/25/2012
From:	Bellamy R R NRC/RGN-I/DRP/PB5
To:	Rich Smith Entergy Nuclear Operations
References
IR-12-002
Download: ML12116A289 (33)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 April 25, 2012 Mr. Robert Smith 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 05000293/2012002

Dear Mr. Smith:

On March 31, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed inspection report documents the inspection results, which were discussed on April 25, 2012 with you and other 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 NRC-identified finding of very low safety significance (Green). This finding was determined to be a violation of NRC requirements. However, because of its very low safety significance, and because it has been entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation (NCV), consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest the non-cited violation in this report, you should provide a written response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-

0001; and the NRC Senior Resident Inspector at PNPS. In addition, if you disagree with the cross-cutting aspect assigned to the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Senior Resident Inspector at PNPS. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Document Access and Managem ent 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/ Ronald R. Bellamy, Chief Reactor Projects Branch 5 Division of Reactor Projects

Docket Nos.: 50-293 License Nos.: DPR-35

Enclosure:

Inspection Report 05000293/2012002

w/Attachment:

Supplemental Information

cc w/encl: Distribution via ListServ

ML12116A289 SUNSI Review Non-Sensitive Sensitive Publicly Available Non-Publicly Available OFFICE RI/DRP RI/DRP RI/DRP NAME MSchneider/TCS for TSetzer RBellamy DATE 04/23/12 04/23/12 04/23/12

1 Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION I

Docket Nos.: 50-293

License Nos.: DPR-35

Report No.: 05000293/2012002

Licensee: Entergy Nuclear Operations, Inc.

Facility: Pilgrim Nuclear Power Station (PNPS)

Location: 600 Rocky Hill Road Plymouth, MA 02360

Dates: January 1, 2012 through March 31, 2012

Inspectors: M. Schneider, Senior Resident Inspector, Division of Reactor Projects (DRP) B. Smith, Resident Inspector, DRP T. Setzer, Senior Project Engineer, DRP K. Dunham, Nuclear Safety Professional Development Program, DRP T. Moslak, Health Physicist, Division of Reactor Safety (DRS)

Approved By: Ronald R. Bellamy, Chief Reactor Projects Branch 5 Division of Reactor Projects

2 Enclosure TABLE OF CONTENTS 1. REACTOR SAFETY .............................................................................................................. 4 1R01 Adverse Weather Protection ...................................................................................... 4 1R04 Equipment Alignment ................................................................................................. 4 1R05 Fire Protection .........................................................................................................

.. 5 1R07 Heat Sink Performance .............................................................................................. 6 1R11 Licensed Operator Requalification Program .............................................................. 6 1R12 Maintenance Effectiveness ........................................................................................ 7 1R13 Maintenance Risk Assessments and Emergent Work Control .................................. 9 1R15 Operability Determinations and Functionality Assessments .................................... 10 1R18 Plant Modifications ................................................................................................... 1 0 1R19 Post-Maintenance Testing ....................................................................................... 11 1R22 Surveillance Testing ................................................................................................ 12 1EP6 Drill Evaluation ......................................................................................................... 12 2. RADIATION SAFETY (RS) .................................................................................................. 13 2RS01 Radioactive Hazard Assessment and Exposure Controls ..................................... 13 2RS02 Occupational ALARA Planning and Controls ........................................................ 15 4. OTHER ACTIVITIES ............................................................................................................ 1 6 4OA1 Performance Indicator (PI) Verification .................................................................... 16 4OA2 Identification and Resolution of Problems ............................................................. 17 4OA3 Follow-Up of Events and Notices of Enforcement Discretion .................................. 17 4OA6 Meetings, Including Exit ........................................................................................... 19 ATTACHMENT: SUPPLEMENTARY INFORMATION................................................................ 19 SUPPLEMENTARY INFORMATION ........................................................................................ A-1 KEY POINTS OF CONTACT .................................................................................................... A-1 LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED ..................................... A-1 LIST OF DOCUMENTS REVIEWED ........................................................................................ A-2 LIST OF ACRONYMS .............................................................................................................

A-11 3 Enclosure

SUMMARY OF FINDINGS

IR 05000293/2012002; 01/01/2012 - 03/31/2012; Pilgrim Nuclear Power Station; Maintenance Effectiveness.

This report covered a three-month period of inspection by the resident inspectors and announced inspections performed by regional inspectors. The inspectors identified one finding of very low safety significance (Green), which was a non-cited violation (NCV).

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). The cross-cutting aspect for the finding was determined using IMC 0310, "Components Within the Cross-Cutting Areas."

Findings for which the SDP does not apply may be Green, or may be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified an NCV of very low safety significance (Green) of 10 CFR Part 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," because Entergy did not include the Rod Worth Minimizer (RWM) system into the scope of Maintenance Rule (MR) systems. Specifically, Entergy did not include the RWM system into the scope of the MR monitoring program as required by 10 CFR 50.65 (b)(2)(i) as a non-safety related system that is relied upon to mitigate accidents or transients. Entergy entered this issue in the corrective action program (CR-PNP-2012-0394).

The inspectors performed a review of IMC 0612, Appendix E, "Examples of Minor Issues," and determined the issue was more than minor because it was similar to example 7.d; in that, the RWM system was not within the scope of the Maintenance Rule and that equipment performance problems were such that effective control of performance could not be demonstrated. The finding was also determined to be more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the availability of the RWM to provide its mitigation function for a control rod drop accident (CRDA). This finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Self Assessment component, because previous assessments performed by Entergy did not include Maintenance Rule scoping attributes nor did they identify scoping issues such as the RWM system. P.3(a) (Section 1R12)

REPORT DETAILS

Summary of Plant Status

Pilgrim Nuclear Power Station began the inspection period during a power ascension to 100 percent reactor power following a December 2011 forced outage. On January 1, Pilgrim returned to 100 percent reactor power. On February 23, operators reduced reactor power to 47 percent to perform a thermal backwash of the main condenser and to perform repairs on main steam isolation valve '2B'. On February 24, Pilgrim returned to 100 percent reactor power. On February 25, operators reduced power to 90 percent to perform a control rod pattern adjustment. Pilgrim returned to 100 percent reactor power later that same day. On March 14, operators reduced power to 65 percent to perform deep and shallow control rod testing, and a subsequent control rod pattern adjustment. Pilgrim returned to 100 percent reactor power later that same day and operated at or near 100 percent reactor power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Readiness for an Impending Storm

a. Inspection Scope

The inspectors performed a review of Pilgrim's readiness for an impending storm on January 12. The review focused on the station's emergency diesel generators, condensate storage tank, and the intake structure. The inspectors reviewed station procedures including Pilgrim's coastal storm and hurricane procedures. The inspectors performed walkdowns of the selected systems to ensure that station personnel had identified issues that could challenge the operability of the systems during high wind conditions. Documents reviewed for each section of this inspection report are listed in

the Attachment.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns (71111.04Q - 3

samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

• 'B' emergency diesel generator with a temporary modification installed on the 24 VDC power supply

• K-110 air compressor with a degraded condition on the K-111 and K-117 air compressors

• Salt service water (SSW) pumps 'C', 'D', and 'E' during 'A' SSW LOOP testing.

The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the Updated Final Safety Analysis Report (UFSAR), technical specifications (TS), work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance or their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the

systems to verify that system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Entergy had properly identified equipment issues and entered them into the corrective action program for resolution with the appropriate significance characterization.

b. Findings

No findings were identified.

.2 Complete System Walkdown

a. Inspection Scope

On January 23 and 24, the inspectors performed a complete system walkdown of accessible portions of the reactor core isolation cooling system to verify the line-up was correct. The inspectors reviewed operating procedures, drawings, equipment line-up check-off lists, and the UFSAR to verify the system was aligned to perform its required safety functions. The inspectors also reviewed electrical power availability, equipment cooling, and hanger and support functionality. The inspectors performed field walkdowns of accessible portions of the system to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. Additionally, the inspectors reviewed a sample of related condition reports and work orders to ensure Entergy appropriately evaluated and resolved any deficiencies.

b. Findings

No findings were identified.

1R05 Fire Protection

Resident Inspector Quarterly Walkdowns (71111.05Q -

5 samples)

a. Inspection Scope

The inspectors conducted tours of the areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that 6 Enclosure Entergy controlled combustible materials and ignition sources in accordance with administrative procedures. The inspectors verified that fire protection identification and suppression equipment was available for use as specified in the area pre-fire plan, and that passive fire barriers were maintained in good material condition. The inspectors also verified that station personnel implemented compensatory measures for out-of-service, degraded, or inoperable fire protection equipment, as applicable, in accordance with procedures.

• Fire Area 1.9, Fire Zone 1.20, Refueling Floor

• Fire Area 1.10, Fire Zone 1.10, Reactor Core Isolation Cooling Pump Quadrant

• Fire Area 1.10, Fire Zone 1.10B, 'B' Residual Heat Removal & High Pressure Coolant Injection Pipe Room

• Fire Area 1.10, Fire Zone 1.28, Reactor Recirculation Pump Motor Generator Sets Room

• Fire Area 1.10, Fire Zone 4.1, 'B' Train Diesel Generator Room

b. Findings

No findings were identified.

1R07 Heat Sink Performance (711111.07A - 1 sample)

a. Inspection Scope

The inspectors reviewed one sample of Entergy's program for maintenance, testing, and monitoring of risk significant heat exchangers (HXs) to assess the capability of the HXs to perform their design functions. The inspectors evaluated whether potential common cause heat sink performance problems could affect multiple HXs in mitigating systems or result in an initiating event. Based on risk significance and performance history, the 'A' Turbine Building Component Cooling Water Heat Exchanger was selected for detailed review by the inspectors.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Requalification Review by Resident Inspectors (1 sample)

a. Inspection Scope

The inspectors observed an emergency drill in the control room simulator on February 29, which included a seismic event followed by an Anticipated Transient Without Scram and then an unisolable leak outside the drywell from the Reactor Water Cleanup system. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the Control 7 Enclosure Room Supervisor. The inspectors verified the accuracy and timeliness of the emergency classifications made by the Shift Manager. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document crew performance problems. Finally, the inspectors performed a simulator fidelity review to determine if the arrangement of the simulator instrumentation, controls, and tagging closely paralleled

that of the control room.

b. Findings

No findings were identified.

.2 Main Control Room Reviews by Resident Inspectors (2 samples)

a. Inspection Scope

The inspectors observed operators perform a thermal backwash of the main condenser

on February 23. Specifically, the ins pectors observed a planned downpower to approximately 47 percent reactor power to support the backwash and maintenance on the main steam isolation valve (MSIV) '2B' DC solenoid valve. The inspectors reviewed procedural guidance for station power changes and the power maneuver plan, and observed Infrequently Performed Test or Evolution briefs. The inspectors observed control room operator performance during the power maneuvers.

The inspectors also observed an infrequently performed evolution on March 14.

Specifically, the inspectors observed a planned downpower to 65 percent to support a control rod pattern adjustment and control rod settle testing. The inspectors reviewed procedural guidance for station power changes and the power maneuver plan, and observed control room operator conduct and control of the evolution.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the samples listed below for items such as:

(1) appropriate work practices;

(2) identifying and addressing common cause failures;

(3) scoping in accordance with 10 CFR 50.65 paragraph

(b) of the Maintenance Rule; (4)characterizing reliability issues for performance;

(5) trending key parameters for condition monitoring;

(6) charging unavailability for performance;

(7) classification and reclassification in accordance with 10 CFR 50.65 paragraphs (a)(1) or (a)(2); and (8)appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as paragraph (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified as paragraph (a)(1).

• 'B' Emergency Diesel Generator Kilowatt Swings

• Neutron Monitoring System

• Rod Worth Minimizer

8 Enclosure

b. Findings

Introduction.

The inspectors identified an NCV of very low safety significance (Green) of 10 CFR Part 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," because Entergy did not include the RWM system into the scope of MR systems. Specifically, Entergy did not include the RWM system into the scope of the MR monitoring program as required by 10 CFR 50.65 (b)(2)(i) as a non-safety related system that is relied upon to mitigate accidents or transients.

Description.

During low power operations, Pilgrim's RWM enforces specific control rod sequences designed to mitigate the consequences of the CRDA by providing control rod blocks to the reactor manual control system. Per Pilgrim's Technical Specification (TS)

Basis document, the RWM enforces the banked position withdrawal sequence below 20 percent reactor thermal power to ensure enthalpy addition due to a potential CRDA does not exceed 280 cal/gm, which is the threshold for fuel damage. Furthermore, Pilgrim's Final Safety Analysis Report (FSAR), Section 14.5, "Postulated Design Basis Accidents,"

assumes the reactor is to be at a control rod pattern corresponding to the maximum incremental worth. It credits a functional RWM in this assumption. The inspectors identified, however, that the RWM was not included within the scope of the MR at Pilgrim. The inspectors also reviewed previous MR assessments, including the prior cycle 10 CFR 50.65 paragraph a(3) assessment, and identified that the evaluation of MR scoping was not an attribute included in these assessments. Additionally, the RWM has had a history of extended unavailability from June to December 2011 and January to March 2012. This history of equipment unavailability does not show effective system performance or condition monitoring through appropriate preventive maintenance activities. Entergy generated CR-PNP-2012-0394 to identify that the RWM system is not scoped in the MR under the consideration of non-safety related systems relied upon to mitigate accidents or transients, utilizing procedure EN-DC-204, "Maintenance Rule Scope and Basis."

Analysis.

The inspectors identified a performance deficiency in that Entergy personnel did not recognize the need to include the RWM system in the MR when it was credited to

mitigate the consequences of a CRDA. Traditional enforcement does not apply since there were no actual safety consequences, impacts on the NRC's ability to perform its regulatory function, or willful aspects of the finding. The inspectors performed a review of IMC 0612, Appendix E, "Examples of Minor Issues," and determined the issue was more than minor because it was similar to example 7.d; in that, the RWM system was not within the scope of the Maintenance Rule and that equipment performance problems were such that effective control of performance could not be demonstrated. The finding was also determined to be more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the availability of the RWM to provide its mitigation function for a CRDA.

The inspectors conducted a Phase 1 screening in accordance with NRC Inspection Manual Chapter (IMC) Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency that did not result in loss of operability, did not represent an actual loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk significant 9 Enclosure per 10 CFR 50.65, and did not screen as risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Self Assessment component, because previous assessments performed by Entergy did not include MR scoping attributes nor did they identify scoping issues such as the RWM system. P.3(a)

Enforcement.

10 CFR Part 50.65, "Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants," requires, in part, that the scope of the monitoring program include non-safety related structures, systems, or components that are relied upon to mitigate accidents or transients. Contrary to the above, Entergy did not scope the RWM's function to mitigate a CRDA into the Maintenance Rule. Entergy's corrective actions included evaluating the RWM for inclusion into the Reactor Manual Control System Maintenance Rule Basis Document. Because this finding is of very low safety significance, and Entergy has entered it into their corrective action program (CR-PNP-2012-0394), this violation is being treated as an NCV, consistent with the NRC's Enforcement Policy. (NCV 05000293/2012002*01, Failure to Scope the Rod Worth Minimizer into the Maintenance Rule)

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed station evaluation and management of plant risk for the maintenance and emergent work activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When Entergy performed emergent work, the inspectors verified that Operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the station's probabilistic risk analyst and Operations personnel to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable

requirements were met.

• Emergent Orange Risk with both the Station Blackout Diesel Generator and the 'B' Emergency Diesel Generator Unavailable

• Green Risk during Planned Condensate Storage Tank Surveillance Testing with Increased Trip Risk

• Green Risk for the Shutdown Transformer being fed from Line 71 and MSIV 2B DC solenoid out-of-service

• Green Risk due to a Degraded Power Supply to the Shutdown Transformer

• Qualitative Yellow Risk Associated with Removing Both Trains of the Standby Gas Treatment System

• Yellow Risk for Analog Trip System Testing

10 Enclosure

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed operability determinations for the following degraded or non-conforming conditions:

• Drywell to Torus differential pressure could not be established while performing Vacuum Breaker Operability Test

• High Pressure Coolant Injection Room Coolers powered off the same AC electrical bus

• Reactor Core Isolation Cooling System Pre-Alarm Low Set Point Actuation

• Reactor Core Isolation Cooling Steam Supply Drain Valve and/or Strainer not Functioning Properly

• Station Blackout Diesel Generator Breaker Failure to Open The inspectors selected these issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and UFSAR to Entergy evaluations to determine

whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled by Entergy. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Modification to Install Temporary Power to the 24 Volt DC (VDC) System

a. Inspection Scope

The inspectors reviewed temporary modification procedure 3.M.3-36.8, Revisions 1 and 2, "Temporary Power for +/- 24VDC Bus 'A' or 'B'", to determine whether the performance capability of the 24VDC system had been degraded through the modification. The inspectors reviewed Control Room drawings, relevant condition reports, and work orders to ensure the temporary modification did not adversely affect the 24VDC system. The inspectors reviewed the annotated drawings to determine whether they properly reflected the temporary modification. The inspectors also walked 11 Enclosure down the temporary power supply in the upper switchgear room to review the installed modification.

b. Findings

No findings were identified.

.2 Permanent Modifications

a. Inspection Scope

The inspectors evaluated a modification to Pilgrim's four Safety Relief Valves and subsequent Reactor Core Isolation Cooling (RCIC) controller calibration implemented by engineering change package EC5000071989, Rev. 8, "SRV/SSV Setpoint & Tolerance Increase and Replacement." The inspectors verified that the design bases, licensing bases, and performance capability of the affected systems were not degraded by the modification. In addition, the inspectors reviewed modification documents associated with the design change, including the design package and RCIC demonstration test procedure 8.5.5.13, Rev. 0, "RCIC Pump Quarterly Operability Flow Rate and Valve Test with Supplemental Pump/Turbine Performance Testing for Benchmarking EC5000071989 Design Assumptions."

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the post-maintenance tests for the maintenance activities listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedures to verify that the procedures adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure was consistent with the information in the applicable licensing basis and/or design basis documents, and that the procedure had been properly reviewed and approved. The inspectors also reviewed test data to verify that the test results adequately demonstrated restoration of the affected safety functions.

• 'B' Emergency Diesel Generator KW Swings and Troubleshooting

• Main Steam Safety Relief Valve SV-203-3D Replacement

• Motor Replacement and Testing for the Residual Heat Removal 'A' Pump

• Repairs to the Rod Worth Minimizer

• Replacement of the DC Coil on the Solenoid to the Main Steam Isolation Valve AO-203-2B

• Replacement of the 'B' Emergency Diesel Generator Motor Operated Potentiometer

• Station Blackout Emergency Diesel Generator Troubleshooting

• Troubleshooting and Repair of a 125VDC Ground on the 'B' Emergency Diesel Generator 12 Enclosure

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk-significant structures, systems, and components (SSCs) to assess whether test results satisfied TS, the UFSAR, and Entergy's procedure requirements. The inspectors verified that test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions. The inspectors reviewed the following surveillance tests.

• 'B' and 'D' Residual Heat Removal Pump Quarterly In-service Testing (IST)

• Emergency Diesel Generator 'A' Initiation by Core Spray Logic

• Main Steam Containment Isolation Valve (CIV) Operability Test

• Reactor Coolant System (RCS) Leakage Detection Surveillance

• Reactor Core Isolation Cooling Valve IST

• Station Blackout Diesel Generator Surveillance

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness)

1EP6 Drill Evaluation

Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of an emergency drill on February 29 to identify any weaknesses and deficiencies in the classification and notification of emergency action levels. The inspectors observed emergency response operations in the simulator and emergency operations facility to determine whether the event classification and notifications were performed in accordance with procedures. The inspectors also attended the station drill critique to compare inspector observations with those identified by Entergy staff in order to evaluate Entergy's critique and to verify whether Entergy's staff was properly identifying weaknesses and entering them into the corrective action program.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Occupational and Public Radiation Safety

2RS0 1 Radioactive Hazard Assessment and Exposure Controls

a. Inspection Scope

During the period of February 13 - 16, the inspector conducted the following activities to verify that Entergy was evaluating, monitoring, and controlling radiological hazards for work performed in locked high radiation areas (LHRA) and other radiological controlled areas, and that workers were adhering to these controls when working in these areas.

Implementation of these controls was reviewed against the criteria contained in 10 CFR Part 20, Technical Specifications, and Entergy's procedures.

Radiological Hazards Control and Work Coverage

The inspector identified exposure significant work areas. The inspector reviewed radiation survey maps and radiation work permits (RWPs) associated with these areas to determine if the associated controls were acceptable. The inspector attended the pre-job and RWP briefings and interviewed selected workers to determine if the workers were informed of the radiological conditions at the job site, electronic dosimeter alarm set points, and actions to be taken if a dosimeter alarms. Specific work activities observed included a High Pressure Coolant Injection (HPCI) pump surveillance test during full power operations.

The inspector toured the accessible radiological controlled area, including the reactor building, turbine building, and radwaste building, and performed independent surveys of selected areas to confirm the accuracy of survey data and the adequacy of postings.

During this tour, the inspector verified that selected LHRAs were properly secured and posted.

In evaluating the RWPs, the inspector reviewed electronic dosimeter dose/dose rate alarm set points to determine if the set points were consistent with the survey indications and plant policy. The inspector verified that workers were knowledgeable of the actions to be taken when the dosimeter alarms, or malfunctions, for tasks being performed under selected RWPs.

The inspector reviewed Entergy's procedure for measuring personnel exposure using the effective dose equivalent method. The inspector confirmed that the method was approved by the NRC and that the implementi ng procedure appropriately specified the placement of whole body and extremity dosimeters on the worker.

14 Enclosure Problem Identification and Resolution

A review of Nuclear Oversight (NO) field observation reports, related Condition Reports (CRs), and an audit report (No.QA-14/15-2011-PNP-1) was conducted to determine if identified problems and negative performance trends were entered into the corrective action program and evaluated for resolution.

Relevant CRs, associated with radiation protection control access, initiated between October 2011 and February 2012, were reviewed and discussed with Entergy staff to determine if the follow up activities were being conducted in an effective and timely manner, commensurate with their safety significance.

High Radiation Area and Very High Radiation Area Controls Procedures for controlling access to High Radiation Areas (HRA) and Very High Radiation Areas (VHRA) were reviewed to determine if the administrative and physical controls were adequate. The inspector also reviewed the physical and procedural controls for securing and removing highly contaminated/activated materials stored in the spent fuel pool. The inspector discussed with radiation protection management the adequacy of current LHRA/VHRA controls, including prerequisite communications and authorizations, and verified that any changes made to relevant procedures did not substantially reduce the effectiveness and level of worker protection. The inspector conducted an inventory of LHRA/VHRA keys to verify accountability.

Radiation Worker Performance and Radiation Protection Technician Performance The inspector observed and questioned radiation workers and radiation protection technicians regarding radiological controls applied to various tasks, including a HPCI test run and various maintenance tasks. The inspector determined that the workers were aware of current RWP requirements, radiological conditions, access controls, and that the skill level was appropriate with respect to the potential radiological hazards and the work being performed.

The inspector attended a Radiation Protection Department daily planning meeting to assess the level of detail provided to workers regarding planned work activities.

The inspector reviewed CRs related to radiation worker and radiation protection technician errors, and personnel contamination event reports to determine if an observable pattern traceable to a similar cause was evident.

Contamination and Radioactive Material Control At the radiological controlled area (RCA) control point, the inspector observed workers surveying and releasing potentially contaminated materials for unrestricted use. The inspector verified that the counting instrumentation was located in a low background area and that the instruments' sensitivity was appropriate for the type of contamination being measured.

The inspector reviewed Entergy's procedures for storing, issuing, and inventorying sealed radioactive sources. The inspector toured plant areas where sealed sources were stored. During this tour, the inspector conducted spot checks of various 15 Enclosure radioactive sealed sources to verify their presence. Through this review, the inspector determined that sources were properly tested for possible leaks, all sources were appropriately accounted for, and that the storage containers were properly secured and labeled.

b. Findings

No findings were identified.

2RS0 2 Occupational ALARA Planning and Controls

a. Inspection Scope

During the period of February 13 - 16, the inspector conducted the following activities to verify that Entergy was properly implementing operational, engineering, and administrative controls to maintain personnel exposure As Low as Reasonably Achievable (ALARA) for tasks performed during 2011 and in performing ongoing activities. Implementation of these controls was reviewed against the criteria contained in 10 CFR Part 20, applicable industry standards, and Entergy's procedures.

Radiological Work Planning

The inspector reviewed pertinent exposure information regarding the spring 2011 refueling outage (RFO), current exposure trends, and ongoing activities to assess ALARA performance. A review of 2011 outage dose was conducted to compare actual exposures with forecasted estimates to determine if differences were properly addressed in Work-In-Progress and Post-Job ALARA reviews.

The inspector evaluated the departmental interfaces between radiation protection, operations, maintenance crafts, and engineering to identify missing ALARA program elements and interface problems. The evaluation was accomplished by attending an ALARA Manager's Meeting for an entry into a steam affected area to troubleshoot a main steam isolation valve (MSIV-203-2B); reviewing the refueling outage post-job ALARA reviews, reviewing the post-job ALARA review for replacing a residual heat removal motor, and reviewing past ALARA Manager's Committee meeting minutes, and NO audit/field observations, and interviewing the site Radiation Protection Manager.

Verification of Dose Estimates

The inspector reviewed the assumptions and basis for the annual (2012) site collective dose, exposure projections and actual exposure data for the 2011 spring outage, and for routine power operations. The inspector evaluated in detail projects whose dose exceeded five person-rem. The inspector reviewed the effectiveness of initial job planning measures and Entergy's efforts in monitoring and controlling dose during job completion by the ALARA Manager's Committee.

The inspector reviewed Entergy's procedures associated with monitoring and re-evaluating dose estimates when the forecasted cumulative exposure for tasks differed from the actual exposure received. The inspector reviewed the dose/dose rate alarm reports, post-job ALARA reviews, and exposure data for selected individuals receiving the highest Total Effective Dose Equivalent (TEDE) for 2011, to confirm that no individual exposure exceeded the regulatory limit or met the performance indicator reporting guideline.

Jobs-In-Progress

The inspector observed a job-in-progress to evaluate the effectiveness of dose and contamination control measures. The job observed was a surveillance test run of a HPCI pump during full power operations. As part of this evaluation, the inspector attended the pre-job and RWP briefing, reviewed the RWP and associated survey maps, and evaluated contamination control measures.

Declared Pregnant Workers

The inspector reviewed the implementing procedure for processing, monitoring, and limiting the exposure of personnel who are Declared Pregnant Workers (DPW). The inspector determined that no DPWs were recently employed who were assigned to perform work within the RCA.

Problem Identification and Resolution The inspector reviewed elements of Entergy's corrective action program related to implementing ALARA program controls, including condition reports, NO field observation reports, audits, and dose/dose rate alarm reports, to determine if problems were being entered at a conservative threshold and resolved in a timely manner.

b. Findings

No findings were identified.

4OA1 Performance Indicator (PI) Verification

Cornerstone:

Initiating Events

a. Inspection Scope

The inspectors reviewed Performance indicator (PI) data to determine the accuracy and completeness of the reported data. The review was accomplished by comparing reported PI data to confirmatory plant records and data available in plant logs, Condition Reports (CRs), Licensee Event Reports (LERs), and NRC Inspection reports. The acceptance criteria used for the review was Nuclear Energy Institute (NEI) 99-02, Revision 6, "Regulatory Assessment Performance Indicator Guidelines." The following performance indicators were reviewed.

• Unplanned SCRAMs per 7000 Critical Hours

• Unplanned SCRAMs with Complications

• Unplanned Power Changes per 7000 Critical Hours

17 Enclosure

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

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

a. Inspection Scope

The inspectors performed a screening of each item entered into Entergy's corrective action program. This review was accomplished by reviewing printouts of each condition report, attending daily screening meetings and/or accessing Entergy'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 were identified.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion

.1 Replacement of 'A' Residual Heat Removal (RHR) Pump Motor

a. Inspection Scope

The inspectors observed operator performance during a replacement of the 'A' RHR pump motor from January 11-17. Specifically, the inspectors reviewed the schedule, replacement activity, TS, and operator actions. The inspectors reviewed procedural guidance for barrier control and observed Entergy's conduct and control of the evolution.

b. Findings

No findings were identified.

.2 Operator Response to a Loss of a DC Solenoid Valve on Main Steam Isolation Valve (MSIV) 2B

a. Inspection Scope

On February 7, Entergy identified that the DC Solenoid Valve for MSIV 2B was inoperable. Two solenoid valves (AC and DC) maintain the MSIVs open and the loss of one of these solenoid valves does not cause an MSIV to close. Entergy evaluated plant risk and established risk management actions to protect the AC solenoid, its power supply, and applicable system relays. Entergy subsequently repaired the DC solenoid during a thermal backwash activity on February 23, and restored the MSIV 2B DC solenoid to service.

b. Findings

No findings were identified.

.3 Operator Performance During Power Maneuvers to Support Main Condenser Thermal

Backwash

a. Inspection Scope

The inspectors observed operators perform a thermal backwash of the main condenser

on February 23. Specifically, the ins pectors observed a planned downpower to approximately 47 percent reactor power to support the backwash and maintenance on the MSIV 2B DC solenoid valve. The inspectors reviewed procedural guidance for station power changes, the power maneuver plan, and observed the Infrequently Performed Test or Evolution brief. The inspectors observed just-in-time training prior to

the evolution and then subsequently observed control room operator performance during the power maneuvers.

b. Findings

No findings were identified.

.4 Operator Performance During a Downpower to Support a Control Rod Pattern

Adjustment

a. Inspection Scope

The inspectors observed an infrequently perform ed evolution on March 14. Specifically, the inspectors observed a planned downpower to support a control rod pattern adjustment and control rod settle testing. The inspectors reviewed procedural guidance for station power changes and the power maneuver plan, and observed control room operator conduct and control of the evolution.

b. Findings

No findings were identified.

.5 (Closed) Licensee Event Report (LER 05000293/2011-005-00), Technical Specification

Required Shutdown Due To Inoperable Feedwater Check Valve The inspectors reviewed Entergy's actions and reportability criteria associated with LER 05000293/2011-005-00, which is addressed in CR-PNP-2011-5228. On November 17, 2011, Pilgrim entered an unplanned shutdown to repair a bonnet pressure seal leak on

the feedwater system check valve 6-CK-62B. Following inspection of the leak, a series of score marks were identified. The valve was repaired and the plant returned to power on November 26, 2011. This LER is closed.

.6 (Closed) Licensee Event Report (LER 05000293/2011-006-00), HPCI Turbine Governor

Control Valve Failure The inspectors reviewed Entergy's actions and reportability criteria associated with LER 05000293/2011-006-00, which is addressed in CR-PNP-2011-5474. On November 30, 2011, Pilgrim declared the High Pressure Coolant Injection (HPCI)system inoperable due to the failure of the turbine governor control valve to open during post-maintenance testing. Following troubleshooting, Entergy determined that the remote servo mechanism had stuck in the "up" position due to mechanical binding. The remote servo mechanism was replaced and the HPCI system was restored to service on December 3, 2011. This LER is closed.

4OA6 Meetings, Including Exit

On February 16, the inspectors conducted a radiation protection exit meeting and presented the results to Mr. Jack Priest, Radiation Protection Manager. At the exit meeting, the inspector confirmed that no proprietary information was provided to the inspector.

On March 29, Ronald Bellamy, NRC Branch Chief for Pilgrim, presented and discussed the 2011 end-of-cycle performance assessment of the Pilgrim Nuclear Power Station with Mr. Robert Smith, Site Vice President, and other members of the Pilgrim staff. The licensee acknowledged the assessment and planned regulatory oversight. This discussion was completed prior to a public open-house meeting on March 29.

(ADAMS Accession ML12074A083).

On April 25, the resident inspectors conducted an exit meeting and presented the preliminary inspection results to Mr. Robert Smith, and other members of the Pilgrim staff. The inspectors confirmed that proprie tary information provided or examined during the inspection was controlled and/or returned to Entergy, and the content of this report includes no proprietary information.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

G. Blankenbiller Chemistry Manager

G. Bradley Component Engineering D. Brugman Supervisor, ALARA/Technical Support B. Chenard System Engineering Manager

B. Clow Radiation Protection Technician

S. Colburn Supervisor Access Authorization and Fitness for Duty

J. Cox Supervisor, Radiation Protection Operations

J. Dent Plant General Manager A. Dodds Director, Nuclear Safety Assurance

V. Fallacara Engineering Director

A. Felix Auxiliary Operator

J. Fitzsimmons Radiation Protection Supervisor J. House Superintendent, Initial Operations Training W. Lobo Licensing Engineer

J. Lynch Licensing Manager

J. Macdonald Assistant Operations Manager-Shift

T. McElhinney Training Manager

W. Mauro Supervisor, Radiation Protection Support

A. Muse Superintendent, Operations Training D. Noyes Operations Manager

J. Priest Radiation Protection Manager

R. Smith Site Vice President

J. Taormina Maintenance Manager M. Thornhill Radiation Protection Supervisor J. Whalley Operations Shift Manager

T. White Emergency Planning Manager

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened and Closed

NCV

05000293/2012002-01 Failure to Scope the Rod Worth Minimizer into the

Maintenance Rule (Section 1R12)

Closed

LER

05000293/2011005-00 Technical Specification Required Shutdown Due To Inoperable Feedwater Check Valve (Section 40A3)

LER

05000293/2011006-00 HPCI Turbine Governor Control Valve Failure (Section 4OA3)

Attachment

Section 1R01: Adverse Weather Protection

Procedures

2.1.37, Coastal Storm Preparations and Actions, Revision 28

5.2.2, High Winds (Hurricane), Revision 31

Miscellaneous

FSAR Section 2.4.4, Storm Flooding Protection

FSAR Section 8.5, Standby AC Power Source

Section 1R04: Equipment Alignment

Procedures

2.2.22, Reactor Core Isolation Cooling (RCIC) System, Revision 72

2.2.22.5, RCIC Injection and Pressure Control, Revision 15 2.2.32, Salt Service Water System, Revision 85 2.2.36, Instrument Air Systems, Revision 71

2.2.37, Service Air System, Revision 33

2.2.8, Standby AC Power System (Diesel Generators), Revision 99

8.C.13, Locked Component Line-up Surveillance, Revision 83

Condition Reports:

CR-PNP-2012-0786, Small air leak on flange of valve 31-HO-607

CR-PNP-2012-1016, Functionality assessment not performed for air leak on 31-HO-607

Miscellaneous

Compensatory Log Sheet for 3/12/2012

FSAR Section 4.7, Reactor Core Isolation Cooling System

RCIC Maintenance Rule Scoping Document

P&ID, FSAR Figure 4.7-1, Drawing M245

P&ID, FSAR Figure 4.7-2, Drawing M246

System Health Report, RCIC, Q3, 2011

Training Manual Documents

Section 1R05: Fire Protection

Procedures

5.5.2, Special Fire Procedure, Revision 49 8.B.17.2, Inspection of Fire Damper Assemblies, Revision 14

8.B.4.11, Fire Panel C225, Reactor Building 23ft., Functional Test, Revision 8

Condition Reports

CR-PNP-2012-0212, NRC Fire Hazard Analysis Review identifies errors

Attachment Miscellaneous Appendix R Safe Shutdown Analysis Report

BECO Letter, May 17, 1983 to the NRC, Appendix R Exemption requests and NRC response to BECO Letter Drawing A316 Sheet 1, Reactor and Turbine, Revision 6

Fire Hazards Analysis Fire Protection Engineering Evaluation (FPEE) 24, RRMG Set Room Ceiling Fire Dampers, Revision 2 2010 and 2011 completed functional tests for smoke detector on RB117

Section 1R07: Heat Sink Performance

Procedures

3.M.4-99, TBCCW HX Tube, Channel Cover, Channel Shell, and Partition Plate

Repair, Revision 17 8.5.3.14, SSW Flow Rate Operability Test, Revision 32

Condition Reports

CR-PNP-2012-0996, Identified a pit on TBCCW HX which requires evaluation

CR-PNP-2012-1009, Flaired ends of TBCCW HX Sleeves broke off during removal

CR-PNP-2012-1062, Several TBCCW HX 'A' tubes have rejectable eddy current results

Miscellaneous

Engineering Change (EC) 27120, New Tube Sleeve Design for RBCCW and TBCCW

Heat Exchangers FSAR, Chapter 10.6, Turbine Building Closed Cooling Water System Heat Exchanger Program Health Report, 4

th Quarter 2011 Operational Decision Making Issue Plan: TBCCW HX High Flow from SSW System Salt Service Water Maintenance Rule Basis Document

Salt Service Water System Health Report, 1

st Quarter 2012 Specification M591, SSW & RBCCW Safety-Related Piping and Heat Exchanger

Inspection, Maintenance & Test Requirements in Response to Generic Letter 89-13 TBCCW Maintenance Rule Basis Document

TBCCW System Health Report, 1

st Quarter 2012

Section 1R11: Licensed Operator Requalification Program

Procedures

2.1.14, Station Power Changes, Revision 107

2.2.94.5, Main Condenser Backwash, Revision 6

8.M.1-8.1, Turbine Control Valve Fast Closure, Revision 0

Miscellaneous

Emergency Action Level Guidelines Instructor Lesson Plan for Thermal Backwash, LP#:0-RQ-04-04-28

Pilgrim Controller Manual Combined Functional Drill (12-01), February 29, 2012

Thermal Backwash Power Profile for 2/23/12 and Power Maneuver Plan

Attachment

Section 1R12: Maintenance Effectiveness

Procedures

EN-DC-203, Maintenance Rule Program, Revision 1

EN-DC-204, Maintenance Rule Scope and Basis, Revision 2

EN-DC-205, Maintenance Rule Monitoring, Revision 3

EN-DC-206, Maintenance Rule (a)(1) Process, Revision 1

2.2.69, Traversing In-Core Probe System, Revision 19

3.M.2-5.6.13, Manual Operation of TIP System, Revision 1

Condition Reports

CR-PNP-2012-0394, Rod Worth Minimizer Not Scoped in the Maintenance Rule

CR-PNP-2011-4261, Rod Worth Minimizer incorrectly set rod position during power ascension

CR-PNP-2011-5799, EDG 'B' KW Swings identified during testing but not in excess of abort criteria

CR-PNP-2012-0043, EDG 'B' KW Swings in Excess of Abort Criteria

CR-PNP-2012-1200, Neutron Monitoring System Not Scoped in the Maintenance Rule

CR-PNP-2010-3705, Rod Block Monitor "A" Downscale

CR-PNP-2012-0048, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-5888, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-4002, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-5739, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-5757, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-5372, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-0794, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-0782, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-0781, Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-1532, Maintenance Rule Functional Failure Evaluation

CR-PNP-2010-3872, Maintenance Rule Functional Failure Evaluation

CR-PNP-2012-1022, LPRM High Alarm Received

CR-PNP-2011-2564, Half Scram on Channel "A"

CR-PNP-2005-2433, #2 TIPS Ball Valve Failure to Open

Miscellaneous

Non Cited Violation (NCV) 05000293/20070201, Inadequate Evaluation of Unexpected Emergency Diesel Generator Load Swings PNPS FSAR Section 14.5, Postulated Design Basis Accidents

PNPS FSAR Section 7.5, Neutron Monitoring System

PNPS FSAR Section 1.6.22, Neutron Monitoring System

PNPS FSAR Section 1.6.4.1.3, Neutron Monitoring System

PNPS FSAR Table 7.5-1, SRM Trips PNPS FSAR Table 5.2-4, Containment and Reactor Vessel Isolation Valves System Health Report, System 45A, Q1-2012

Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 2 Technical Specifications and Bases, Section 3.3.F, Rod Worth Minimizer Vermont Yankee Scope and Basis Document for the Rod Worth Minimizer Maintenance Rule SSC Basis Document, Neutron Monitoring System, Revision 1 Maintenance Rule SSC Basis Document, Primary Containment Isolation System, Revision 2

Maintenance Rule SSC Basis Document, Primary Containment System, Revision 1

Attachment

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

1.5.22, Risk Assessment Process, Revision 15 2.2.80, Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation, Revision 47

2.4.A.23, Loss/Degradation of 23KV Line

8.M.1-32.7, Analog Trip System - Trip Unit Calibration Cabinet C2233B Section A, Revision 39

8.M.1-32.8, Analog Trip System Trip Unit Calibration Cabinet C2233B Section B, Revision 30

8.M.2-2.5.6, HPCI Condensate Storage Tank Level, Revision 35 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument

Calibration, Revision 48

EN-WM-104, On Line Risk Assessment, Revision 6

Condition Reports

CR-PNP-2011-3791, Decision Made to Remove Both Trains of the Standby Gas Treatment System

CR-PNP-2011-5950, During Diesel Pump Post Maintenance Test, seal found to be leaking

CR-PNP-2012-0657, SDT will be powered from an alternate source from 2/6/2012 through

4/17/2012.

Risk and EOOS will be updated to reflect this condition

CR-PNP-2012-0737, Risk hammocks for DC solenoid were not reflected in weekly schedule

CR-PNP-2012-1358, Standby Gas Treatment Test Procedure Inaccurate

Miscellaneous

Control Room Logs

Equipment Out-Of-Service Risk Assessment Tool Drawing MIJ16-10, Elementary Diagram HPCI System, Revision E25 Drawing E730, Elementary Diagram ECCS Analog Trip Cabinet C2233B, Revision E7

Final Safety Analysis Report Chapter 5.3, Secondary Containment System Risk Assessment Profile for 1/4/2012

Risk Profiles for Weeks 2/13/2012 and 2/20/2012 Technical Specification 3.7.B, Standby Gas Treatment System and Control Room High

Efficiency Air Filtration System

Section 1R15: Operability Determinations and Functionality Assessments

Procedures

ARP-EPIC, Alarm Response Procedure, Revision 34 2.2.70, Primary Containment Atmospheric Control System, Revision 109

2.2.146, Station Blackout Diesel Generator, Revision 42

2.4.143, Shutdown from outside the control room, Revision 48

5.3.31, Station Blackout, Revision 16 8.A.1, Drywell to Torus Vacuum Breaker Monthly/Quarterly Operability, Revision 47 8.5.5.1, RCIC Pump Quarterly and Biennial Operability Flow Rate and Valve Test, Revision 74

8.9.16.1, Manually Start and Load the Blackout Diesel from the Shutdown Transformer, Revision 42

Condition Reports

CR-PNP-2007-4857, RCIC Flow Controller downscale

CR-PNP-2010-4252,

LV-2301-31 has developed packing leak

CR-PNP-2011-5534,

LV-2301-31 has small steam leak

CR-PNP-2012-0375, RCIC Pre-Alarm Low Set point Actuation Attachment

CR-PNP-2012-0715, Drywell to Torus dp could not be established

CR-PNP-2012-0052, Station Blackout Diesel Breaker Failure to Open

CR-PNP-2012-1251, RCIC Steam Supply Drain Pot Valve blockage

CR-PNP-2012-1352, During troubleshooting RCIC

CV-1301-32 was identified to have a body to bonnet steam leak

Miscellaneous

Calculation N120, HPCI Pump Room Heatup without unit coolers, Revision 0

Control Room Logs for 2/11/2012 Drawing E10, Single Line Diagram - 480V System Motor Control Centers B14, B15, B17,

B18, B28, & B29, Revision 53

Drawing M-245, P&ID RCIC System, Revision E35

Final Safety Analysis Report Chapter 8.10, Blackout AC Power Source

NRC Open Item 91-04.01.1 Design Adequacy of HPCI Area Coolers Station Blackout Diesel Maintenance Rule Basis Document

Section 1R18: Plant Modifications

Procedures

2.2.15, 24V DC Battery System, Revision 5 3.M.3-36.8, Temporary Power for +/- 24VDC Bus 'A' or 'B', Revisions 1 and 2

8.5.5.1, RCIC Pump Quarterly and Biennial Operability Flow Rate and Valve Test at Approximately 1000 PSIG, Revision 74 8.5.5.13, RCIC Pump Quarterly Operability Flow Rate and Valve Test with Supplemental Pump/ Turbine Performance Testing for Benchmarking EC5000071989 Design Assumptions, Revision 0

EN-DC-136, Temporary Modifications, Revision 6

EN-OM-119, On-site Safety Review Committee, Revision 8

Condition Reports

CR-PNP-2011-4228, No Post-Installation Electrical Check for Installing the +/- 24VDC Temporary Modification

CR-PNP-2012-1136, Temporary Modification Tag Installed on Incorrect Drawing

CR-PNP-2012-1151, Out-of-Date Reference listed on 24VDC Alarm Response Procedure

Miscellaneous

DC Power System Health Report for 1

st Quarter 2012 Engineering Change EC5000071989, SRV/SSV Setpoint & Tolerance Increase & Replacement, Revision 8

Final Safety Analysis Report, Chapter 8.7, 24VDC Power System Neutron Monitoring Maintenance Rule Basis Document 10

CFR 50.59 Evaluation of Procedural Temporary Modification 3.M.3.36.8 dated 7/24/2009 24VDC Maintenance Rule Basis Document Technical Specifications 3.5.D, Reactor Core Isolation Cooling System

Section 1R19: Post-Maintenance Testing

Procedures

2.1.31, Rod Worth Minimizer Operability, Revision 13

3.M.3-51, Electrical Termination Procedure, Revision 28

Attachment 3.M.3-5.10, 4KV

POWL-VAC Vacuum Circuit Breakers 152-801 and 152-802 Maintenance and Inspection 3.M.3-61.7, EDG Woodward Governor Tuning Procedure, Revision 10 3.M.4-8.1, Main Steam Isolation Valve Preventive Maintenance, Revision 14 8.5.2.2.1, LPCI Loop 'A' Operability Test, Revision 54

8.5.6.2, Special Test for ADS Syst em Manual Opening of Relief Valves

8.7.4.4, Main Steam Isolation Valve Operability 60% Power, Revision 24

8.9.1, Emergency Diesel Generator and Associated Emergency Bus Surveillance, Revision 123

8.9.96.1, Manually start and load the Blackout Diesel via the Shutdown Transformer, Revision

EN-WM-107, Post Maintenance Testing, Revision 3

EN-MA-125, Troubleshooting Control of Maintenance Activities, Revision 8

Temporary Procedure (TP)12-002, Test for EDG 'B' Governor Replacement, Revision 0

Condition Reports

CR-PNP-2012-0109, During PWT of Motor Operated Potentiometer Replacement 'B' EDG

exhibited 200-250KW Swings

CR-PNP-2012-0700, Discrepancies identified during review of SRV Work Order Packages

CR-PNP-2012-0043, 'B' EDG KW Swings during Testing

CR-PNP-2012-0070, Received "'B' 125VDC Ground Alarm"

CR-PNP-2012-0784, Missing Post Maintenance Test for 125VDC Ground Troubleshooting

CR-PNP-2012-0821, Post Maintenance Test Form for SBO Diesel Troubleshooting

CR-PNP-2012-0079, Station Blackout Replacement Failed Acceptance Criteria

CR-PNP-2012-0052, SBO Diesel Breaker A801 would not open during testing

CR-PNP-2012-0125, Loose wiring indentified in terminal J220

CR-PNP-2012-0127, EDG 'B' Frequency did not stabilize following governor testing

CR-PNP-2012-1277, Additional PMT required for 125VDC troubleshooting performed on the 'B'

EDG

Maintenance Orders/Work Orders

WO 00183050 08, Replace EDG 'B' Governor

WO 00183050 09, Inspect/Repair Connections in J220

WO 00183050 04, Replace EDG 'B' Governor, Post Maintenance Test

WO 00183050 01, Replace EDG 'B' Mechanical Governor

WO 00299451 10, Replace

SRV-203-3C Next Cold Shutdown (CR-PNP-2011-5388), OPS PMT

WO 00301212 02, Exercise 'B' EDG Woodward Governor Motor Operated Potentiometer (MOP)

WO 00301212 03, Post Work Test 'B' EDG (MOP)

WO 00301212 08, EDG 'B' Inoperable due to Kilowatt Load Swings, EDG 'B' Troubleshoot for

25VDC Ground

WO 00301212 11, EDG 'B' Troubleshooting

WO 00300689 03, Received Alarm-Relief/Safety Valve Leaking for

SV-203-3D, Ops Post Work Test

WO 00300962, Task 1, Attachment 1, RWM Operate Startup from Cold Shutdown

WO 00305324, Tasks 2, 5, & 7, Troubleshoot apparent failed coil,

SV-203-20-1

WO 00301251 01, Perform 3.M.3-5.10 on 152-801

WO 00301251 02, Troubleshoot installed 152-801 Breaker per

EN-MA-125

WO 00301251 03, Perform Ops Post Work Testing (8.9.16.1)

WO 00301251 04, Remove Breaker 152-801 and perform inspection per V1096

WO 00309220, Task 1, Perform Troubleshooting PMT on EDG 'B' Control Circuit

WO 51526849, Task 1, Clean P-203A Screens & Openings Attachment

WO 52339009, Task 1, 8.Q.3-2, EQ Insulation Test (P-203A & 5KV Cable)

WO 52193640, Task 1, Calibrate Sensors on RHR P-203A

Miscellaneous

Control Room Logs

PM List for RHR Motors Vendor Manual V-0367, GE Induction Motors, Revision 11

Section 1R22: Surveillance Testing

Procedures

2.5.2.71, Radwaste Collection System, Revision 32

8.5.2.2.2, LPCI System Loop 'B' Operability - Pump Quarterly Flow Rate Test, Revision 46

8.5.5.4, RCIC Motor Operated Valve Quarterly Operability Test, Revision 39

8.7.4.4, Main Steam Isolation Valve Operability, 60% Power, Revision 24

8.9.16.1, Manually Start and Load Blackout Diesel via the Shutdown Transformer

8.M.2-2.10.8.3, Diesel Generator 'A' Initiation by Core Spray Logic

Condition Reports

CR-PNP-2011-5743, Drywell Floor Sump HI and LO Alarms not working

CR-PNP-2012-0155, RCIC Valve

AO-1301-34 failed to stroke close

Maintenance Orders/Work Orders

WO 52385124, Task 1, LPCI Loop 'B' Pump Quarterly Operability

WO 52370387, Task 1, Quarterly MSIV

Technical Specifications Technical Specifications (TS), Sections 3.7 and 3.13

Section 1EP6: Drill Evaluation

Miscellaneous

Pilgrim Controller Manual Combined Functional Drill (12-01), February 29, 2012

Emergency Action Level Guidelines Sections 2RS01 and SRS02: Radiological Hazard Assessment/ALARA Planning &

Controls

Procedures

EN-RP-101, Access Control for Radiologically Controlled Areas

EN-RP-105, Radiological Work Permits

EN-RP-108, Radiation Protection Posting

EN-RP-122, Alpha Monitoring

EN-RP-131, Air Sampling

EN-RP-143, Source Control

EN-RP-204, Special Monitoring Requirements

PNP-RP-6.1-22, Radiological Controls for High Risk Evolutions

Attachment

Condition Reports

2011-5808

2012-0199

2011-5124

2012-6963 2011-4554

2012-0221 2011-5434

2011-5651

2011-5342

2012-0249

2011-5817

2012-0795

2012-0159

Radiation Work Permits (RWP) and associated ALARA Plans

2012-085, Replace 'A' RHR Pump Motor

2012-083, Thermex Upgrade

2012-064, Thermex High Impact Critical Points Identification

2012-062, Walkdown in Locked High Radiation Areas (LHRA) and High Radiation Areas 2012-052, Instrumentation & Control Activities in LHRA 2012-089, Troubleshoot and Repair MSIV

SV-203-2B-1 DC

ALARA Post Job Review

RWP 2012-085, Replace RHR Pump (P-203A) motor and support activities Nuclear Oversight Audit and Field Observations

CR-PNP-2011-3544

CA 15, Contamination Control Process

LO-WTPNP-2010-0277, Alpha Monitoring

LO-PNPLO-2011-0085, Radiation Protection Program

QA-14/15-2011-PNP-1, Radiation Protection and Radwaste

QS-2011-PNPS-012, Radiation Protection Document Control

Miscellaneous

Reports

Dose and Dose Rate Alarm Report for period 10/01/2011 through 02/13/2012

Dose Report for the Ten Highest Personnel Exposure for 2011 5-Year Dose Reduction Plan Dose Reduction Initiatives for 2012

Section 4OA1: Performance Indicator Verification

Procedures

EN-LI-114, Performance Indicator Process, Revision 5

Condition Reports

CR-PNP-2012-0181, Incorrect PI data on NRC Web site

Miscellaneous

NRC Inspection Report 1

st Quarter 2011 through 4

th Quarter 2011 NRC PI Web site Pilgrim NRC Performance Indicator Technique/Data Sheet

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion

Procedures

2.1.14, Station Power Changes, Revision 107

2.2.92, Main Steam Line Isolation and Turbine Bypass Valves, Revision 53

Attachment

2.2.94.5, Main Condenser Backwash, Revision 6

2.4.30, MSIV Closure, Revision 20

8.C.42, Subcompartment Barrier Control Surveillance, Revision 23

8.M.1-8.1, Turbine Control Valve Fast Closure, Revision 0

EN-HU-101, Attachment 9.7, Station Consequential Error, Yellow Sheet 1/16/132, Incorrect Configuration of Sandbags Challenged Flooding Analysis, Revision 8

EN-LI-102, Corrective Action Process, Revision 17

Condition Reports

CR-PNP-2011-5228, Feedwater Check Valve 6-CK-62B Pressure Seal Leak and Associated Root Cause Report

CR-PNP-2012-0190, A RHR Pump Trips in Torus

CR-PNP-2012-0289, Lower Motor Bearing Temp Reading Erratic

CR-PNP-2012-0288, Results of Baker Testing Low

CR-PNP-2012-0231, Mispositioned Sandbags

CR-PNP-2012-0655, MSIV 2B DC Solenoid Failure

CR-PNP-2012-1335, NRC Identified in Feedwater Check Valve Root Cause, CRs for package closeout were not evaluated in Cause

Miscellaneous

Instructor Lesson Plan for Thermal Backwash, LP#:0-RQ-04-04-28

NUREG 1022, Event Report Guidelines 10CFR 50.72 & 50.73, Revision 3

Main Steam Drawings Power Maneuver Plan Thermal Backwash Power Profile for 2/23/12 and Power Maneuver Plan

Attachment

LIST OF ACRONYMS

ADAMS Agencywide Documents Access and Management System

ALARA As Low As Reasonably Achievable

CA corrective action

CFR Code of Federal Regulations

CR condition report

CRDA control rod drop accident

DRP Division of Reactor Projects

DRS Division of Reactor Safety

EDG emergency diesel generator

EP emergency preparedness

FSAR Final Safety Analysis Report

HPCI high pressure coolant injection

HRA high radiation area

HX heat exchangers

IMC Inspection Manual Chapter

ISI in-service inspection

LHRA locked high radiation area

MR maintenance rule

MSIV main steam isolation valve

NO nuclear oversight

NCV non-cited violation

NRC Nuclear Regulatory Commission

PI performance indicator

PNPS Pilgrim Nuclear Power Station

RCA radiologically controlled area

RCIC Reactor Core Isolation Cooling

RHR residual heat removal

RWM rod worth minimizer RWP's radiation work permit(s)

SDP significance determination process

SSC structure, system, or component

SSW salt service water

TS technical specification

UFSAR Updated Final Safety Analysis Report

VHRA very high radiation areas

WO work order