IR 05000293/2011004

IR 05000293-11-004; on 07/01/2011-09/30/2011; Pilgrim Nuclear Power Station; Flood Protection Measures, Maintenance Effectiveness, and Maintenance Risk Assessment and Emergent Work Control

ML11306A054
Person / Time
Site:	Pilgrim
Issue date:	11/02/2011
From:	Bellamy R R NRC/RGN-I/DRP/PB5
To:	Rich Smith Entergy Nuclear Operations
LHP
References
IR-11-004
Download: ML11306A054 (40)

Text

UNITED STATES NUCLEAR REGU LATORY GOMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA. PENNSYLVANIA 19406.1415 November 2, ?OLL Mr. Robert Smith Site Vice President Entergy Nuclear Operations, lnc.Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508

SUBJECT: PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/201 1004

Dear Mr. Smith:

On September 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed inspection report documents the results, which were discussed on October 13,2011, 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 three NRC-identified findings of very low safety significance (Green).These findings were determined to be violations of NRC requirements.

However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC's Enforcement Policy. lf you contest any NCV, you should provide a 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 l; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident lnspector at PNPS. ln addition, if you disagree with the cross-cutting aspect assigned to any 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 l, and the NRC Senior Resident Inspector at PNPS. The information you provide will be considered in accordance with lnspection Manual Chapter 0305.ln accordance with 10 CFR 2.390 of the NRC's "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 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,Ronald R. Bellamy, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-293 License No. DPR-35

Enclosure:

InspectionReport05000293/2011004 M

Attachment:

Supplemental lnformation cc: Mencl: Distribution via ListServ

SUMMARY OF FINDINGS

lR 0500029312011004 07fi112011-0913012011;

Pilgrim Nuclear Power Station; Flood Protection Measures, Maintenance Effectiveness, and Maintenance Risk Assessment and Emergent Work Control.The report covered a three-month period of inspection by the resident and regional-based inspectors.

Three non-cited violations (NCVs) of very low safety significance (Green) were identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using lnspection Manual Chapter (lMC) 0609, "significance Determination Process." Cross-cutting aspects associated with findings are determined using IMC 0310, "Components Within the Cioss-Cutting Areas." 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.NRC-ldentified Findinqs Gornerstone:

Mitigating Systems

Green.

The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion lll, Design Control, because Entergy's design control measures did not ensure two-over-one seismic protection of the 'C' Salt Service Water (SSW) Pump. Specifically, Entergy did not ensure that a Class I to Class ll interface would not result in a failure of a Class I component ('C'SSW Pump). Corrective actions included installing a temporary modification (i.e., water shield), to protect the pump motor from potential spray effects of a Class ll piping failure and performing an extent of condition review.The inspectors performed a review of Inspection Manual Chapter (lMC) 0612, Appendix E, "Examples of Minor lssues," and did not find a similar more than minor example. The finding was determined to be more than minor because it was associated with the Protection Against External Events (i.e., seismic) attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone's objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the 'C' SSW pump motor was vulnerable to water spray from a failed Class ll pipe during a seismic event which could have rendered the pump inoperable.

The inspectors used IMC 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," and determined that further evaluation was required since the finding was potentially risk significant due to a seismic initiating event. As a result of this screening, a Phase 3 evaluation was conducted by a regional Senior Reactor Analyst (SRA). The condition was assessed as Green, with a change in core damage frequency (CDF)calculated to be 1.29E-8. Since the finding was assessed to have a CDF of less than 1E-7,large early release frequency was not required to be assessed.

The finding does not have a cross-cutting aspect since the failure to verify the adequacy of design with respect to ensuring two-over-one seismic protection for the 'C' SSW pump is not indicative of current licensee performance.

In addition, current Entergy design procedures require rigorous Class ll-over-l criteria for all new modifications. (Section 1 R06)

4 Gornerstone:

Barrier IntegritY

Green.

The inspectors identified a Green NCV of 10 CFR 50.65, paragraph (a)(1) and (a)(4, "Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power planis," because Entergy did not monitor the performance of the Primary Containment System (Drywell to Torus Vacuum Breaker Components)against license-established goals to provide reasonable assurance that these components are capable of fulfilling iheir intended functions.

Specifically, Entergy did not identify a functional failure of the Drywell to Torus Vacuum Breaker Component portion of the Primary Containment Sy-stem and thereby did not recognize that the system exceeded its unavailability performance criteria, requiring a Maintenance Rule (aX1) evaluation.

Entergy subsequently conducted an (aX1) evaluation and concluded that the system should be classified as (a)(1), corrective actions specified, and system monitoring completed.

The finding is more than minor because it is associated with the Barrier Performance attribute of the Barrier Integrity cornerstone, in that the issue affected the Primary Containment System reliability due to the failure to recognize the need to evaluate the system for goais, corrective actions, and monitoring.

The inspectors determined the significance of the finding using IMC 0609-04, "Phase 1 - Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) because the degraded condition had been corrected by the time of the failure to accurately evaluate the maintenance rule functionalfailure.

As a result, this finding did not involve a design or qualification deficiency, did not result in a loss of systeft safety function, and did not screen as potentially risk significant due to external initiating events. The finding has a cross-cutting aspect in the Human Performance cross-cutting area, Decision Making component; in that, Entergy did not use conservative assumptions when evaluating the degraded Drywell to Torus Vacuum Breakers condition to correctly conclude that a functional failure had occurred'Specifically, Entergy did not consider that the function of these vacuum breakers would be requireb as soon as plant conditions exceeded 212F, and therefore, the procedural guidance for Technical Specification applicability not being exceeded was an incorrect basis for this decision H.1(b). (Section 1 R12)Green. The inspectors identified a Green NCV of 10 CFR 50.65(aX4)because Entergy did not assess and manage risk during elective maintenance for both 'A' and 'B' trains of the StandBy Gas Treatment (SBGT) system. Specifically, Entergy did not consult qualitative guidance in their risk assessment process procedures before removing both tiains of SgCt from service and, therefore, removing the Secondary Containment key safety function while online. Corrective actions planned include evaluating and revising risk assessment procedures, and communicating qualitative risk assessment guidance to Senior Reactor Operators and Work Week Managers.A review of NRC Inspection Manual Chapter (lMC) 0612, Appendix E, "Minor Examples," identified that Section 7, Maintenance Rule, Example e, reflected a similar more than minor example. This finding was determined to be more than minor because Entergy's risk assessment failed to account for the loss or significant uncompensated impairment of a key operating safety function.

In addition, the finding affected the Human performance attribute of the Barrier Integrity cornerstone's objective to ensure that physical design barriers (containment)protect the public from radionuclide releases caused by accidents or events. The inspectors performed an evaluation in accordance with IMC 0609, "significance Determination Process," Attachment 4, "Phase 1 -lnitial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the SBGT system. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Entergy did not plan work activities by incorporating appropriate risk insights H.3(a). (Section 1R13)

.1 a.6

REPORT DETAILS

Summarv of Plant Status Pilgrim Nuclear Power Station (PNPS) began the inspection period operating at 100 percent reactor power. On July 21, operators reduced power to 50 percent reactor power to perform a thermal backwash on the main condenser.

Pilgrim returned to 100 percent reactor power on July 22. On July 25, operators reduced power to 90 percent reactor power to perform a control rod pattern adjustment and returned to 100 percent reactor power later that same day. On Sepiember 20, operators reduced power to 50 percent reactor power to perform a thermal backwash on the main condenser.

Pilgrim returned to 100 percent reactor power on September 21, and operated at or near 100 percent reactor power for the remainder of the inspection period.1. REACTOR SAFEW Cornerstones:

Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather Protection (71111.01 - 2 samples)External Floodinq Inspection Scope During the week of August 15, the inspectors reviewed Pilgrim's plant design for coping with the design basis probable maximum flood. The inspectors reviewed the "Storm Flooding Protection" section of the Updated Final Safety Analysis Report and operating procedures for mitigating externalflooding conditions during severe weather. The inspectors also performed a walkdown of the site to determine if all susceptible flooding conditions had been considered in the plant design, and whether operating procedures could be reasonably carried out to mitigate flooding concerns.

Documents reviewed for each section of this inspection report are listed in the Attachment.

Findinqs No findings were identified.

lmpendinq Storm Inspection Scope During the week of August 22, Hurricane lrene was tracking to impact the Pilgrim Plant over the weekend. The inspectors reviewed Entergy's preparations for the hurricane and the high winds expected to accompany the storm. The inspectors also performed a walkdown of the outside areas including the switchyard to determine if loose debris or other material could become airborne in the presence of high winds and thereby impact safety related equipment.

As the hurricane moved through the region, inspectors were staffed at the site continuously.

The inspectors verified the availability of systems important to safety by monitoring conditions and alarms in the control room and technical support center. The inspectors verified that operator actions defined in Entergy's b.a..2 Enclosure 7 adverse weather procedure maintained the readiness of essential systems. The inspectors discussed readiness and staff availability for adverse weather response with operations and work control personnel and monitored Entergy's contingency staffing of emergency response facilities.

The inspectors conducted site walkdowns after winds had abated to ensure no adverse conditions arose from this storm. Documents reviewed during this inspection are listed in the Attachment.

b. Findinqs No findings were identified.

1R04 Equipment

Alignment (71111.04)

.1 Partial Svstem Walkdowns

(71111.04Q - 3 samples)a. Inspection Scope The inspectors performed three partial system walkdowns during this inspection period.The inspectors performed a partial walkdown of each system to determine if the critical portions of the selected systems were correctly aligned in accordance with procedures, and to identify any discrepancies that may have had an effect on operability.

The walkdowns included selected control switch position verifications, valve position checks, and verification of electrical power to critical components.

In addition, 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:

o 'A' Residual Heat Removal during 'B' Core Spray Header Differential Pressure Test r 'A' Emergency Diesel Generator during a maintenance window on the Station Blackout (SBO) Diesel Generator o Reactor Core lsolation Cooling during a maintenance window on the SBO Diesel Generator and Shutdown Transformer b. Findinqs No findings were identified.

.2 Complete Svstem Walkdowns

(71111.04S - 1 sample)a. Inspection Scope The inspectors completed a detailed review of the High Pressure Coolant Injection (HPCI) system to assess the functional capability of the system. The inspectors performed a walkdown of the system to determine whether the critical components, such as valves and breakers were aligned in accordance with operating procedures, and to assess the material condition of valves and other supporting equipment.

The inspectors discussed system health with the system engineer, reviewed the system's Maintenance Rule status, and performed a review of outstanding maintenance work orders to determine whether the deficiencies significantly affected the HPCI system function.

The Enclosure 8 inspectors also reviewed condition reports from the past year to determine whether HPCI equipment problems were being identified and appropriately resolved.

The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R05 Fire Protection

(71111.05)

.1 Annual Fire Drill Observation

(71111.05A - 1 sample)a. Inspection Scope The inspectors observed an announced fire drill in the'A'4160VAC Switchgear Room.The drill was conducted in accordance with procedure EN-DC-189, Revision 1 , "Fire Drills." The inspectors observed performance of the fire brigade personnel to determine whether Entergy's fire fighting pre-plan strategies were utilized, the pre-planned drill scenario was followed, and the drill objectives were met. The inspectors confirmed that protective clothing and breathing apparatus were donned; sufficient firefighting equipment was brought to the scene; the fire brigade leader's fire fighting directions were clear; and communications with the plant operators and between fire brigade members were effective.

The inspectors observed the drill critique to determine whether areas to improve fire brigade performance were identified.

Findinqs No findings were identified.

Fire Protection - Tours (71111.05Q - 5 samples)Inspection Scope The inspectors performed walkdowns of five fire protection areas during the inspection period. The inspectors reviewed Entergy's fire protection program to determine the 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 and fire barriers.

The inspectors then compared the existing condition of the areas to the fire protection program requirements to determine whether all program requirements were met. The documents reviewed during this inspection are listed in the Attachment.

The fire protection areas reviewed were: o Fire Area 4.3, Fire Zone 4.3,'A' Emergency Diesel Generator Room. Fire Area 1.9, Fire Zone 2.3, 'A' Battery Room r Fire Area 1.10, Fire Zone 2.4, 'B' Battery Room. Fire Area 5.3, Fire Zone 5.6, Electric Fire Pump Area and Open Areas of the lntake Structure b.a..2 Enclosure

.1 I. Fire Area 5.3, Fire Zone 5.4, Diesel Driven Fire Pump Room

b. Findinqs No findings were identified.

1R06 Flood Protection

Measures (71111.06 - 1 sample)Internal Floodinq Inspection Inspection Scope The inspectors walked down the intake structure including the Salt Service Water compartments, Sea Water pump rooms, Diesel Fire pump and fueltank rooms, and associated flood propagation pathways to assess the effectiveness of Entergy's internal ftood control measures.

The inspectors assessed the condition of curbing and selected flood pathways.

The inspectors also evaluated whether potential sources of internal flooding were analyzed.Findinqs lntroduction.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion lll, Design Control, because Entergy's design control measures did not ensure two-over-one seismic protection of the 'C' Salt Service Water (SSW) Pump. Specifically, Entergy did not ensure that a Class I to Class ll interface would not result in a failure of a Class I component

('C' SSW Pump).Description.

The inspectors reviewed potential internalflooding sources affecting safety-related equipment in Pilgrim's Intake structure.

The inspectors identified a potential vulnerability in the 'C' SSW pump cubicle in that Class ll city water piping carrying lubricating and motor bearing cooling water to the circulating water pumps is housed adjacent to the "C" SSW pump motor. The inspectors discussed this with Entergy's design engineering department to determine if there was a potential flooding scenario that tould affect the safety-related equipment.

Entergy walked down the area and concluded that the condition had not been previously analyzed.

Entergy generated CR-PNP-201 1-3729 and determined that the 'C' SSW motor could be susceptible to direct spray impingement from the Class ll city water piping during certain seismic event scenarios.

Entergy reviewed vendor specifications and consulted with the vendor concluding that although the motor is a weather proof, drip proof design, it is not designed for direct spray impingement.

Pilgrim's Updated Final Safety Analysis Report (UFSAR) Section 12.2.3.5, "Seismic Loids," discusses design criteria for seismic loading including criteria concerning Class ll/Class I interfaces.

lt states "Class I to Class ll interfaces are designed so that there will be no functional failure in the Class I structure.

In order to accomplish this design objective, Class I structures have the capacity of withstanding the forces resulting from possible failures of Class ll structures which are either attached or adjacent to the Class i Structures." Pilgrim's TDBD-118, Revision E0, "Design Basis Document for Seismic Loading", clarifies design expectations further and adds that "since about 1983, rigorous Class ll-over-l criteria have been applied to all station modifications.

This action, b.Enclosure 10 combined with the consideration and resolution of seismic interaction hazards provides reasonable assurance that the UFSAR requirement is met'" Following the determination that the 'C' SSW pump motor would be susceptible to direct spray from Class ll piping during a seismic event, Pilgrim declared the'C' SSW inoperable per their Technical Specifications and developed a temporary modification thai installed a "shield" for protection from this flooding scenario.

Pilgrim also conducted extent-of-condition walkdowns around the plant for other potential Class ll/Class I spray concerns and found none.Analvsis.

The inspectors determined that the failure to verify the adequacy of design with respect to ensuring adequate two-over-one seismic protection for the 'C' SSW pump was a performance deficiency within Entergy's ability to foresee and correct and should have been prevented.

Specifically, Pilgrim's UFSAR and seismic design basis documents specify, in part, that the failure of a class ll structure will not cause a failure of a class I structure.

ln addition, design control measures for verifying the adequacy such as a dynamic analysis or verification checks had not been performed pertaining to this vulneribility.

This condition did not impact the regulatory process and did not contribute to any actual consequences; therefore, Traditional Enforcement did not apply. The inspectors performed a review of Inspection Manual Chapter (lMC) 0612, Appendix E,"Examples of Minor lssues," and did not find a similar more than minor example to apply.The finding was determined to be more than minor because it was associated with the Protection Against External Events (i.e. seismic) attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone's objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the 'C' SSW pump motor was vulnerable to spray during a seismic event that could have rendered the pump inoperable.

The inspectors used IMC 0609.04,"Phase 1 - Initial Screening and Characterization of Findings," and determined that further evaluation was required since the finding was potentially risk significant due to a seismic initiating event. As a result of this screening, a Phase 3 evaluation was conducted by a regional Senior Reactor Analyst (SRA).The SSW system cools the Reactor Building Closed Cooling Water system and has five motor driven pumps. The system is configured with train 'A' composed of pumps 'A'and'B', train 'B' composed of pumps 'D'and 'E', and pump 'C' being a swing pump. The SSW system is normally cross-tied with a total of at least two pumps running. The 'C'pump can be aligned to either train 'A' or'B'. The condition was assessed as a seismically induced transient.

The exposure period was assumed to be 1 year. lt was also assumed that for all measured seismic events the 'C' SSW pump would fail due to water impingement.

The seismic transient frequency of 1E-2lyr was developed from the Pilgrim Individual Plant Examination for External Events (IPEEE). No recovery of the 'C'SSW pump was assumed. Based on these assumptions the condition was assessed as Green, with a change in core damage frequency calculated to be 1

.29 E-8. Since the finding was assessed to have a CDF of less than 1E-7,large

early release frequency was not required to be assessed.The finding does not have a cross-cutting aspect since the failure to verify the adequacy of design with respect to ensuring adequate two-over-one seismic protection for the 'C'SSW pump is not indicative of current licensee performance.

ln addition, current Enclosure 11 Entergy design procedures require rigorous Class ll-over-l criteria for all new modifications.

Enforcement.

10 CFR 50, Appendix B, Criterion lll, Design Control, requires, in part, that measures be provided for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, since initial plant design, Entergy's design control measures had not verified that the adequacy of design with respect to ensuring adequate two-over-one seismic protection existed for the'C' SSW pump. Specifically, Entergy had not performed a design review to ensure the 'C' SSW pump would not be affected from Class ll piping during a seismic event. Corrective actions included installing a temporary modification (i.e., water shield), to protect the pump motor from potential spray effects of the Class ll piping failure and performing an extent of condition review. Because this violation was of very low safety significance and was entered into Entergy's corrective action program (CR-PNP-2O11-3729), this violation is being treated as an NCV, consistent with the NRC's Enforcement Policy. (NCV 05000293/2011004-01, Failure to Verify the Adequacy of the Design for the 'G' Salt Service Water Pump).

1R07 Heat Sink Performance

(71111.07 - 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 assessed whether the HX program conformed to Entergy's commitments at Pilgrim related to NRC Generic Letter 89-13,"Service Water System Problems Affecting Safety-Related Equipment." ln addition, 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 its risk significance and performance history, the 'A' Reactor Building Closed Cooling Water HX was selected for a detailed review by the inspectors.

b. Findinqs No findings were identified.

1R1 1 Licensed Operator Requalification Program (71111.11)

Resident Inspector Quarterlv Review (71111

.1 1 Q - 1 sample)a. Inspection

Scope The inspectors observed licensed operator performance during an emergency preparedness drill on September

7. The inspectors

observed crew response to a Hostile Action Based scenario which resulted in the loss of the onsite Fire Protection System and the Intake Structure.

The inspectors assessed the licensed operators'performance to determine if the drill evaluators adequately addressed observed deficiencies during the post-drill critique.

The inspectors also reviewed the applicable drill objectives from the scenario to determine if they had been achieved.

ln addition, the inspectors b.

1R12 12 performed

a simulator fidelity review to determine if the arrangement of the simulator instrumentation, controls, and tagging closely paralleled that of the control room.Findinqs No findings were identified.

Maintenance Effectiveness (71 1 I 1.12Q).1 Equipment Failure Evaluations (4 samples)a. Inspection Scope The inspectors reviewed the four 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 paragraph (aX1 ) or (aX2); 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 (aX1).. Control Room Envelope Functional Failure Evaluation. Drywell to Torus Vacuum Breaker Functional Failure Evaluation. HPCI Drain Valves Functional Failure Evaluation. 'B' Reactor Building Closed Cooling Water Heat Exchanger Functional Failure Evaluation b. Findinqs lntroduction.

The inspectors identified an NCV of very low safety significance (Green) of I O Cfn 50.65 paragraph (aX1) and (a)(2), "Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants," because Entergy did not monitor the performance of the Primary Containment System (Drywell to Torus Vacuum Breaker Components)against license-established goals to provide reasonable assurance that these components are capable of fulfilling their intended functions.

Specifically, Entergy did not identify a functional failure of the Drywell to Torus Vacuum Breaker Component portion of the Primary Containment System and thereby did not recognize that the system exceeded its unavailability performance criteria requiring a Maintenance Rule (aX1) evaluation.

The subsequent evaluation concluded that the system should be classified as (a)(1), corrective actions specified, and system monitoring completed.

Description.

On May 13, Entergy was unable to establish the required differential pressure netween the drywell and the torus (suppression chamber)during plant start-up.Entergy performed a plant shutdown and determined that several Drywell to Torus Vacuum Breakers were leaking by, which precluded the ability to establish the differential pressure.

On May 26, System Engineering evaluated the condition and concluded that the issue was not a functional failure since the plant was not operated Enclosure 13 beyond the point at which the drywell to torus differential pressure was required to be established by plant technical specifications.

The inspectors subsequently reviewed the basis for this conclusion and determined that the function(s)of the Drywell to Torus Vacuum Breakers would have been required independent of this specific technical specification (i.e., when the susceptibility to a Loss of Coolant Accident above 212F was established during plant heat-up).

Entergy re-evaluated the condition and concluded that a maintenance preventable functionalfailure had occurred since a maintenance activity conducted during the refueling outage had incorrectly adjusted several vacuum breakers and post work testing was not completed to identify the leaking condition prior to plant start-up.

As a result of the re-classification of this degraded condition, System Engineering evaluated the status of the Primary Containment System and concluded that the system should be classified under 10 CRF50.65(a)(1), goals and corrective actions established, and system monitoring specified.

Analvsis.

The inspectors determined that Entergy's failure to identify the Dry,vell to Torus Vacuum Breakers condition as a functionalfailure, and as a result, the failure to perform an evaluation of the system under 50.65(a)(1)and thereby specify goals, corrective actions, and monitoring, was a performance deficiency within Entergy's ability to foresee and correct and should have been prevented.

Traditional Enforcement did not apply, as the issue did not have actual or potential safety consequence, had no willful aspects, and did not impact the NRC's ability to perform its regulatory function.

A review of NRC Inspection Manual Chapter (lMC) 0612, Appendix E, "Minor Examples," revealed that no minor examples were applicable to this finding. The finding is more than minor because it is associated with the Barrier Performance attribute of the Barrier Integrity cornerstone, in that, the issue affected the Primary Containment System reliability due to the failure to recognize the need to evaluate the system for goals, corrective actions, and monitoring.

The inspectors determined the significance of the finding using IMC 0609-04, "Phase 1 - lnitial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green)because the degraded condition had been corrected by the time of the failure to accurately evaluate the maintenance rule functional failure. As a result, this finding did not involve a design or qualification deficiency, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events.The finding has a cross-cutting aspect in the Human Performance cross-cutting area, Decision Making component; in that, Entergy did not use conservative assumptions during the evaluation of the degraded Drywell to Torus Vacuum Breakers to correctly conclude that a functional failure had occurred.

Specifically, Entergy did not consider that the function of these vacuum breakers would be required as soon as plant conditions exceeded 212F and therefore the procedural guidance for Technical Specification applicability not being exceeded was an incorrect basis for this decision.lH.1(b)l

Enforcement.

10 CFR 50.65 (aX1), requires, in part, that the holders of an operating license shall monitor the performance or condition of structures, systems, or components (SSCs) within the scope of the rule as defined by 10 CFR 50.65 (b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions.

10 CFR 50.65 (aX2) states, in part, that monitoring as specified in 10 CFR 50.65 (aX1) is not required where it has Enclosure 14 been demonstrated that the performance or condition of an SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function.

Contrary to the above, on May 26, Entergy incorrectly evaluated a Primary Containment System component failure which precluded System Engineering from evaluating the system for 10 CFR 50.65(aX1)monitoring requirements.

Subsequently, Entergy re-evaluated the Primary Containment System for this functional failure and determined that monitoring under 10 CFR 50.65(a)(1)would be required.Corrective actions taken for this violation included revising the Maintenance Rule functional failure evaluation for this equipment, classifying the Primary Containment System as a 10 CFR 50.65(a)(1)system, and specifying goals, corrective actions, and monitoring for the system. Because this violation was of very low safety significance and was entered into Entergy's corrective action program (CR-PNP-2011-2993 and -3210), this violation is being treated as an NCV, consistent with the NRC's Enforcement Policy. (NCV 05000293/2011004-02, Failure to ldentify a Primary Containment System Maintenance Rule Functional Failure and Thereby Establish Monitoring Requirements for the System)..2 Maintenance Rule (aX3) Evaluation Review (1 sample)a. Inspection Scope The inspectors performed a review of the Entergy assessment of the Pilgrim Maintenance Rule program implementation as specified by 10 CFR 50.65(a)(3).

The inspectors evaluated whether this assessment was conducted within the periodicity required by 10 CFR 50.65(aX3).

The inspectors also evaluated whether Entergy reviewed 10 CFR 50.65(a)(1)goals and 10 CFR 50.65(a)(2)performance criteria.Preventive maintenance and corrective action effectiveness associated with this program were also reviewed.

In addition, the inspectors evaluated the use of lndustry Operating Experience within the program and whether adjustments to the program were made as a result of the periodic assessment.

The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R13 Maintenance

Risk Assessments and Emergent Work Control (71111

.13 - 4 samples)a. Inspection

Scope The inspectors evaluated four maintenance risk assessments for emergent and planned testing and maintenance activities.

The inspectors reviewed maintenance risk evaluations, work schedules, and control room logs to determine if concurrent maintenance or surveillance activities adversely affected the plant risk already incurred with out-of-service components.

The inspectors evaluated whether Entergy took the necessary steps to controlwork activities, minimized the probability of initiating events, and maintained the functional capability of mitigating systems. The inspectors assessed Entergy's risk management actions during plant walkdowns.

The inspectors reviewed Enclosure 15 the conduct and adequacy of maintenance risk assessments for the following maintenance and testing activities:. Green Risk during Load Shed Testing of the 'A' and 'C' Residual Heat Removal and 'A'Core Spray Pumps. Green Risk during 'A' Emergency Diesel Generator Load Shed Testing. Yellow Risk during maintenance on the Station Blackout Diesel Generator and the Shutdown Transformer and an emergent issue with offsite power Line 342. Green Risk during maintenance on both 'A' and 'B' trains of the Standby Gas Treatment system b. Findinqs

Introduction.

The inspectors identified a Green NCV of 10 CFR 50.65(a)(4)because Entergy did not assess and manage risk during elective maintenance for both the 'A'and'B' trains of the StandBy Gas Treatment (SBGT) system. Specifically, Entergy did not consult qualitative guidance in their risk assessment procedure before removing both trains of SBGT, thereby removing the Secondary Containment key safety function while online.Description.

On August 3, Pilgrim elected to perform maintenance on both the 'A' and'B' SBGT demister drain valves, rendering both trains of SBGT unavailable.

Pilgrim's Secondary Containment System (SCS) is designed to be sufficiently leaktight to allow at least one train of SBGT to reduce reactor building pressure to a minimum sub-atmospheric pressure of 0.25 inches of water and for SBGT to treat assumed leakage rates and fission products entrapped in the SCS. Another function of the SCS is to limit the ground level release to the environs of airborne radioactive materials so that offsite doses from a design basis fuel accident or loss of coolant accident will be below the guideline values stated in 10 CFR Part 100, "Reactor Site Criteria." Entergy's Equipment Out of Service (EOOS) risk assessment model does not quantitatively model SBGT nor SCS, since the absence of their function does not contribute quantifiably to core damage frequency (CDF). However, Pilgrim's Secondary Containment is a key safety function that prevents or mitigates the consequences of accidents that could result in potentially significant off-site exposures.

Thus, a qualitative risk evaluation for the absence of a key safety function is warranted.

As described in Entergy's procedure EN-WM-104, Revision 4, Online Risk Assessment, the definition of a Qualitative Risk Assessment, in part, is "an evaluation of the risk of maintenance based on judgment, in which a broad spectrum of potential impacts on plant safety and operation are considered.

These may include, but are not limited to, Technical Specifications, defense in depth, impacts on key safety functions, and radiological/AlARA." Furthermore, Pilgrim's procedure 1.5.22, Revision 14, Risk Assessment Procedure, Section 5.4.2, Qualitative Risk Assessment Guidelines states,"Maintenance activities degrading the integrity of Primary and/or Secondary Containment can increase Large Early Release Frequency.

As Primary and/or Secondary Containment is not modeled in EOOS, the risk associated with these activities will be qualitatively assessed by raising the EOOS color one level; e.9., green to yellow." lt goes on to discuss, "The PSA model does not address the radioactivity release protection afforded.

lt is entirely possible that planned SBGT maintenance can degrade the radioactivity release mitigation function." Enclosure 16 The inspectors identified that qualitative considerations were not discussed during the work planning process and that procedure 1.5.22 and EN-WM-104 were not consulted.

Thus, the qualitative aspects of removing the secondary containment key safety function were not evaluated by station personnel.

Entergy entered this issue into their corrective action program as CR-PNP-2011-3791.

Corrective actions planned include evaluating and revising onsite procedure 1.5.22 to better match EN-WM-104 in regard to qualitative criteria as well as to improve proficiency for Senior Reactor Operators and Work Week Managers in the use of qualitative risk assessment guidelines Analvsis.

The performance deficiency associated with this finding is that Entergy did not correctly perform a risk assessment using qualitative criteria as outlined in station procedures for elective maintenance of both trains of SBGT as specified by 10 CFR 50.65(aX4).

The performance deficiency was within Entergy's ability to foresee and correct and should have been prevented.

Traditional Enforcement did not apply as the issue did not have actual or potential safety consequence, had no willful aspects, nor did it impact the NRC's ability to perform its regulatory function.A review of NRC Inspection Manual Chapter (lMC) 0612, Appendix E, "Minor Examples," identified that Section 7, Maintenance Rule, Example e, reflected a similar more than minor example. This finding was determined to be more than minor because Entergy's risk assessment failed to account for the loss or significant uncompensated impairment of a key operating safety function.

In addition, the finding affected the Human Performance attribute of the Barrier Integrity cornerstone's objective to ensure that physical design barriers (containment)protect the public from radionuclide releases caused by accidents or events. The inspectors performed an evaluation in accordance with IMC 0609, "significance Determination Process," Attachment 4, "Phase 1 -lnitial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the SBGT system.The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, work control component, because Entergy did not plan work activities by incorporating appropriate risk insights.

H.3(a)]Enforcement.

10 CFR 50.65 paragraph (aX4), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," States, in part, that "'..the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities." Contrary to the above, on August 3, Energy did not correctly assess the risk of removing the SCS safety function.

As a result, Entergy did not recognize an increased risk condition and thus did not take risk management actions.Corrective actions planned include evaluating and revising onsite procedure 1.5.22to better match EN-WM-104 in regard to qualitative criteria as well as improve proficiency for Senior Reactor Operators and Work Week Managers in the use of qualitative risk assessment guidelines.

Because of the very low safety significance and because it has been entered into the corrective action program (CR-PNP-2011-3791), the NRC is treating this as a non-cited violation (NCV), consistent with Section 2.3.2 a of the NRC's Enforcement Policy. (NCV 05000293/2011004-03, Failure to Accurately Assess Risk of Maintenance on Standby Gas and Secondary Gontainment)17 1R15 Operabilitv Evaluations (71111.15 - 4 samples)a. Inspection Scope The inspectors reviewed four operability determinations associated with degraded or non-conforming conditions to determine if the operability determination was justified and if the mitigating systems or barriers 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 Entergy's performance against related Technical Specifications and UFSAR requirements.

The inspectors reviewed the following degraded or non-conforming conditions:. CR-PNP-2011-3344, Residual Heat Removal Loop 'A' Containment Spray Header Flow Transmitter Power Supply Ripple Voltage Out of Specification. CR-PNP-2011-3424, High Pressure Coolant Injection Turbine Exhaust Line Drain Valves Open when they are Normally Closed. CR-PNP-2011-3733, Failure to Include Seismic Input in Channel-Control Blade lnterference Guidance. CR-PNP-2011-4164 and CR-PNP-2011-4200, 'B'Standby Liquid Control Degraded Conditions b. Findinqs No findings were identified.

1R19 Post-Maintenance

Testinq (71111.19 - 7 samples)a. Inspection Scooe The inspectors reviewed seven samples of post-maintenance tests during this inspection period. The inspectors reviewed these activities to determine whether the post-maintenance test 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.

ln addition, the inspectors evaluated the applicable test acceptance criteria to verify consistency with the associated design and licensing bases, as well as Technical Specification requirements.

The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution.

The following maintenance activities and their post-maintenance tests were evaluated:. Replace the Recirculation Flow Converter providing input to the 'B'Average Power Range Monitor Flow-Biased Scram Setpoint. Replace Internals on the'D' Reactor Building Component Cooling Water Discharge Check Valve. Replace Standby Gas Treatment Demister Drain Valves. Replace Standby Gas Treatment Damper AO-N-98 Enclosure 18. Replace the Drain Valve on the Automatic Depressurization pressure switch for the'B'Residual Heat Removal PumP. Post Installation Test on the Alternate Charger for the 'A' DC 24V Batteries. Preventative Maintenance and Testing on Air Cooled Breaker 103 b. Findinqs No findings were identified.

1R22 Surveillance

Testinq (71111.22 - 5 samples)a. Inspection Scope The inspectors witnessed five surveillance activities and/or reviewed test data 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 Technical Specification requirements.

The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution.

The following surveillance tests were evaluated:. 'D' Reactor Building Closed Cooling Water Pump Biennial Comprehensive In-Service Test (lST)o 'B'Core Spray Pump and Valve Quarterly IST. High Pressure Coolant Injection Cold Quickstart Test. 'A' Salt Service Water Loop Flow Rate Operability Test. 'B' Salt Service Water Pump Quarterly IST b. Findinqs No findings were identified.

Gornerstone:

Emergency Preparedness 1EP2 Alert and Notification Svstem (ANS) Evaluation (71114.02 - 1 sample)Inspection Scope An onsite review was conducted to assess the maintenance and testing of the Pilgrim Nuclear Power Station ANS. During this inspection, the inspectors interviewed EP staff responsible for implementation of the ANS testing and maintenance, and reviewed Condition Reports pertaining to the ANS for causes, trends, and corrective actions. The inspectors reviewed the ANS procedures and the ANS design report to ensure Entergy's compliance with design report commitments for system maintenance and testing. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Enclosure 02. Planning Standard, 10 CFR 50.47(bX5)and the related requirements of 10 CFR 50, Appendix E, were used as reference criteria.b. Findinos No findings were identified.

1EP3 Emerqencv Response Orqanization (ERO) Staffinq and Auqmentation Svstem (71114.03 - 1 sample)a. Inspection Scope The inspectors performed a review of Pilgrim's ERO augmentation staffing requirements and the process for notifying and augmenting the ERO. This was conducted to ensure the readiness of key licensee staff to respond to an emergency event and to ensure Entergy's ability to activate their emergency facilities in a timely manner. The inspectors reviewed the Pilgrim ERO roster, training records, applicable procedures, drill reports for augmentation, quarterly EP drill reports, and CRs related to the ERO staffing augmentation system. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 03. Planning Standard, 10 CFR 50.47(bX2)and related requirements of 10 CFR 50, Appendix E, were used as reference criteria.b. Findinqs No findings were identified.

1EP4 EmerqencvAction Level (EAL) and Emerqencv Plan Chanoes (71114.04 - 1 sample)a. Inspection Scope Since the last NRC inspection of this program area, in November 2010, Entergy had implemented various revisions of the different sections of the Pilgrim Nuclear Power Station Emergency Plan. Entergy had determined that, in accordance with 10 CFR 50.54(q), any change made to the Plan, and its lower-tier implementing procedures, had not resulted in any decrease in effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR 50. The inspectors reviewed all EAL changes that had been made since November 2010, and performed a sampling review of other Emergency Plan changes, including the changes to lower-tier emergency plan implementing procedures and EP-related equipment, to evaluate for any potential decreases in effectiveness of the Emergency Plan. However, this review was not documented in an NRC Safety Evaluation Report and does not constitute formal NRC approval of the changes.Therefore, these changes remain subject to future NRC inspection in their entirety.

The inspection was performed in accordance with NRC Inspection Procedure 71114, Attachment 04. The requirements in 10 CFR 50.5a(q) were used as reference criteria.b. Findinqs No findings were identified.

b.20 1EPs Correction of Emerqencv Preparedness Weaknesses (71114.05 - 1 sample)lnspection Scope The inspectors reviewed a sampling of self-assessment procedures and reports to assess Entergy's ability to evaluate their EP performance and programs.

The inspectors reviewed a sampling of condition reports from January 2010 through July 2011, initiated by Entergy at Pilgrim from drills, self-assessments and audits. Additionally, the inspectors reviewed Quality Assurance audits, including 10 CFR 50.54(t) audits, and several self-assessment reports. This inspection was performed in accordance with NRC lnspection Procedure71114, Attachment 05. Planning Standard, 10 CFR 50.47(b)(14) and the related requirements of 10 CFR 50 Appendix E were used as reference criteria.Findinqs No findings were identified.

lEPO Drill Evaluation (71114.06 - 1 sample)a. Inspection Scope The inspectors observed a licensed operator emergency preparedness drill on September

7. The inspectors

evaluated operator performance in the simulator for a Hostile Action Based scenario which resulted in the loss of the onsite Fire Protection Systems and the Intake Structure.

The scenario escalated from an unusual event to a general emergency.

The inspectors assessed the implementation of Emergency Action Level classification and notification decisions for this event. The inspectors also assessed whether Pilgrim's critique of the exercise assessed all observations and findings.b. Findinqs No findings were identified.

2. RADTATTON

SAFETY (RS)Gornerstones:

Occupational and Public Radiation Safety 2RS08 Radioactive Solid Waste Processinq and Radioactive Material Handling.

Storaqe and Transportation (7 1 I 24.08)a. Inspection Scope During the period August 8 through August 11, the inspector performed the following activities to verify that Entergy effectively implemented their programs for processing, handling, storage and transportation of radioactive material.

lmplementation of these controls was reviewed against the criteria contained in 10 CFR Part20, relevant Technical Specifications, and the licensee's procedures.

21 lnspection Planninq The inspector reviewed the solid waste system description in Pilgrim's Updated Final Safety Analysis Report, Pilgrim's Process Control Program, and Pilgrim's 2010 Annual Effluent Release Report.The inspector reviewed Pilgrim's 2009 audit, QA-1 4115-2009-PNP-01, of the Radiation Protection

/ Radwaste program.Radioactive Material Storaqe The inspector observed the storage of containers of radioactive material in the Trash Compactor Facility (TCF) yard area and other areas of the site. The inspector verified the containers were properly labeled.The inspector verified that the radioactive material storage areas were properly posted and controlled.

The inspector verified that Entergy has established a process for monitoring the impact of long-term storage of radioactive waste.The inspector verified that there were no signs of swelling or leakage of the containers used to store radioactive materials.

Radioactive Waste Svstem Walkdown. The inspector walked down the accessible portions of the liquid and solid radioactive waste systems including the reactor water clean-up system, the chemical waste clean-up system, clean waste clean-up system, and the spent resin processing system.. The inspector verified the concentrator that was abandoned in place is isolated and will not contribute to an unmonitored release path.. The inspector verified there have been no changes to the radioactive waste processing system since the last inspection in 2009'. The inspector verified that the waste stream mixing, sampling procedures, and methodology for waste concentration averaging are consistent with the Process Control Plan (PCP) and provide representative sampling of the spent resin for waste classification.. The inspector verified that the liquid waste tanks for discharge are recirculated to provide sufficient mixing.. The inspector verified the PCP contains references to procedures that correctly describe the current methods for dewatering and waste stabilization.

Waste Characterization and Classification. The inspector reviewed the analyses for two waste streams and verified that Entergy's radiochemical sample analysis results were sufficient to support accurate radioactive waste characterization.

The inspector verified that Entergy's use of scaling factors, dose rates and dose to curie conversion factors, and calculations to account for difficult-to-measure radionuclides is technically sound and based on current 10 CFR Part 61 analyses'. The inspector verified that Entergy's procedures take into account changing plant Enclosure b.22 operational parameters, and additional samples are obtained for 10 CFR Part 61 analyses when needed to maintain the validity of the waste stream composition data between the annual and biennial sample analysis.. The inspector verified that Entergy has established and maintains an adequate Quality Assurance program to ensure compliance with the waste classification and characterization requirements.

Shipment Preparation. The inspector observed the loading of a spent resin liner into a transport cask and torquing of the lid for transport.

The inspector observed the labeling, marking, placarding, vehicle checks, shipping papers provided to the driver, and licensee verification of shipment readiness.

The inspector verified that the receiving licensee was authorized to receive the shipment packages'. The inspector observed radiation protection technicians during the conduct of radioactive waste processing and radioactive material shipment preparation.

The inspector verified that the personnel were knowledgeable of the shipping regulations and demonstrated adequate skills to accomplish the package preparation req u i rements for publ ic transport.

Shippinq Records. The inspector reviewed three Type A shipping packages and verified the documents indicated the proper shipper name; emergency response information including a 24-hour contact telephone number; accurate curie content and volume of material; appropriate waste classification; and UN number.Problem ldentification and Resolution. The inspector reviewed Pilgrim's self-assessments and audits related to the solid radioactive material control program to determine if identified problems were entered into the corrective action program. The inspector verified that problems identified were put into the corrective action program and appropriate corrective actions were identified.

Findinos No findings were identified.

OTHER ACTIVITIES

]OAI Performance Indicator (Pl) Verification (71 151)4.40A1.1 Cornerstone:

Mitiqatinq Svstems (3 samples)a. Inspection Scope The inspectors reviewed Pl data to determine the accuracy and completeness of the reported data. The review was accomplished by comparing reported Pl data to Enclosure b.a..2 23 confirmatory plant records and data available in plant logs, condition reports, Licensee Event Reports, and NRC inspection reports. The acceptance criteria used for the review was Nuclear Energy Institute (NEl) 99-02, Revision 6, "Regulatory Assessment Performance lndicator Guidelines." The following performance indicators were reviewed:. High Pressure Coolant Injection System from the third quarter of 2Q10 through the second quarter of 2011 [MS07]. Heat Removal System from the third quarter of 2Q10 through the second quarter of 2011 [MS08]. Residual Heat Removal System from the third quarter 2010 through the second quarter of 2011 [MS09]Findinqs No findings were identified.

Cornerstone:

Emerqencv Preparedness (EP) (3 samples)Inspection Scope The inspectors reviewed data for the Pilgrim EP Pls, which are:

(1) Drill and Exercise Performance (DEP);

(2) Emergency Response Organization (ERO) Drill Participation; and,

(3) Alert and Notification System (ANS) Reliability.

The last NRC EP inspection at Pilgrim was performed in the fourth quarter of 2010, so the inspectors reviewed supporting documentation from EP drills, training records, and equipment tests from the fourth calendar quarter of 2010 through the second quarter of 2011, to verify the accuracy of the reported Pl data. The review of these Pls was conducted in accordance with NRC lnspection Procedure 71151, using the acceptance criteria documented in NEI gg-02, "Regulatory Assessment Performance Indicator Guidelines," Revision 6.Findinqs No findings were identified.

ldentification and Resolution of Problems (71152)Review of ltems Entered into the Corrective Action Prooram (CAP)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.

Findinqs No findings were identified.

b.4c.42.1 a.b.Enclosure

.1 24 Annual Sample: Operator Workarounds

Inspection Scope The inspectors performed the annual review of operator workarounds to verify Entergy was identifying operator workaround problems at an appropriate threshold and entering them into the corrective action program. The inspectors reviewed identified workarounds to determine whether the mitigating system function was affected, whether the operator's ability to implement abnormal and emergency operating procedures was affected, and whether appropriate procedures had been updated to reflect actual plant conditions.

The inspection was accomplished through personnel interviews, plant tours, and review of station documents.

Findinqs No findings were identified.

Operator workarounds have been identified and entered into the corrective action program for resolution.

No unrecognized impacts to operator or system performance were identified, and corrective actions have been implemented or are planned to restore the affected systems.Event Follow-up (71 153)(2 samples)Inspection Scope The inspectors observed operators perform a condenser backwash and control rod testing on July 21 and September 20. Specifically, the inspectors observed planned plant downpowers to approximately 50 percent reactor power to support backwashes of the main condenser on these dates. The inspectors reviewed procedural guidance for station power changes and the power maneuver plan, and observed the lnfrequently Performed Test or Evolution briefs. The inspectors also observed control room operator performance during the power maneuvers and in response to unexpected plant conditions.

Findinqs No findings were identified.

'A'and 'B'Trains of Salt Service Water (SSW) Svstem Declared Inoperable (1 sample)Inspection Scope On September 22, Entergy identified that all five SSW pumps were susceptible to failure during certain degraded voltage scenarios.

Entergy evaluated plant risk and entered the appropriate Technical Specification (TS) which was to place the reactor in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with both trains of SSW inoperable.

Entergy performed a modification that removed the design vulnerability and exited TS before having to shutdown.

The b..2 a.Enclosure 25 inspectors responded to the control room, reviewed Entergy's actions, plant risk, and TS administration.

b. Findinqs No findings were identified.

.3 (Closed) Licensee Event Report (LER 05000293/201

1-001-00).

Technical Specification (TS) Required Shutdown - Reactor Buildinq Closed Coolinq Water (RBCCW)'B' Declared Inoperable The inspectors reviewed Entergy's actions and reportability criteria associated with LER 05000293/2011-001-00, which is addressed in CR-PNP-2011-0721.

On February 20, Pilgrim commenced a shutdown of the reactor due to the 'B' RBCCW system being declared inoperable and expected to exceed its 72-hour TS Limiting Condition for Operability.

NRC Inspection Report 05000293/2011002, Section

4OA3 documents

the event and the inspectors' response.

Following the event, repair activities identified a single tube leak in the 'B' RBCCW heat exchanger related to a shortened inlet end sleeve. No findings or violations of NRC requirements occurred.

This LER is closed..4 (Closed) Licensee Event Report (LER 05000293/2011-002-00).

Reactor Scram Durinq a Planned Reactor Cool-Down with All Control Rods Fullv lnserted The inspectors reviewed Entergy's actions associated with LER 05000293/2011-002-00, which is addressed in CR-PNP-2011-0733.

On February 20, with the reactor shutdown and all control rods fully inserted, a valid Reactor Protection System low reactor water level scram initiation signal was received.

At the time of the event, a reactor cooldown was in progress and the Reactor Mode Selector Switch was in "Startup".

Entergy performed a causal analysis and determined that the scram actuation signal was the result of reactor water level control difficulties during the cooldown using the Mechanical Pressure Regulator.

Reactor Water level was immediately restored and the scram signal was reset. No findings or violations of NRC requirements occurred.

This LER is closed..5 (Closed) Licensee Event Report (LER 05000293/2011-003-00).

Reactor Scram on lntermediate Ranqe Monitor Hiqh-Hioh Flux The inspectors reviewed Entergy's actions and reportability criteria associated with LER 05000293/2011-003-00, which is addressed in CR-PNP-2011-2475.

On May 10, a reactor scram event occurred at Pilgrim during a reactor plant start-up.

A Special Inspection Team (SlT) was chartered and arrived on-site on May 16. The SIT reviewed the event, interviewed personnel, and reviewed Entergy's root cause analysis.

NRC Inspection Report 05000293/2011012 was issued on September 1, and documents the results of the SIT inspection.

This LER is closed..6 (Closed) Licensee Event Report (LER 05000293/2011-004-00).

Technical Specification (TS) Required Shutdown Drwell to Torus DP The inspectors reviewed Entergy's actions associated with LER 05000293/2011-004-00, which is addressed in CR-PNP-2011-2538.

On May 14, plant operators were unable to Enclosure 26 establish a drywell to torus differential pressure as specified by TS. The plant was shutdown, the problem investigated, and several drywell to Torus Vacuum Breakers were found leaking due to an improper magnet to striker plate clearance.

Entergy performed a causal analysis and determined that the vacuum breakers had been incorrectly adjusted during refueling outage maintenance activities due to insufficient procedural guidance.

The vacuum breakers were re-adjusted and a plant startup was commended.

No findings or violations of NRC requirements occurred.

This LER is closed.4OAO Meetinqs.

Includinq Exit On July 28, an Emergency Preparedness exit meeting was conducted with Mr. Stephen Bethay, Director of Nuclear Safety Assurance, and other members of the Entergy staff.The inspector confirmed that proprietary information was not provided or examined during the inspection.

On August 11, a Radiation Safety exit meeting was conducted with Mr. Vincent Fallacara, Director of Engineering (and Acting Site Vice President).

The inspector confirmed that no proprietary information was provided to the inspector for the inspection.

On October 13, 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 proprietary 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:

=SUPPLEMENTAL

INFORMATION=

KEY POINTS OF CONTACT

Entergy personnel:

S. Brewer Radiation

Protection

Supervisor

D. Brugman Radiation

Protection

Supervisor

B. Chenard System Engineering

Manager J. Dreyfuss Plant General Manager V. Fallacara

Engineering

Director E. Herbert l&C Supervisor

W. Lobo Licensing

Engineer

J. Lynch Director, Nuclear Safety Assurance

and Licensing

Manager J. Macdonald

Assistant

Operations

Manager-Shift

T. McElhinney

Training Manager D. Noyes Operations

Manager M. O'Meara System Engineer R. Pace Design Engineering

Supervisor

J. Priest Radiation

Protection

Manager M. Santos Chemistry

Technician

J. Scheffer Chemistry

Supervisor

K. Sejkora Senior Chemistry

Specialist

R. Smith Site Vice President J. Taormina Maintenance

Manager J. Whalley Operations

Shift Manager T. White Emergency

Planning Manager

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened and Closed

05000293/2011-004-01

NCV Failure to Verify the Adequacy of the Design for the 'C'Salt Service Water PumP

05000293/2011-004-02

NCV Failure to ldentify a Primary Containment

System Maintenance

Rule Functional

Failure and Thereby Establish Monitoring

Requirements

for the System

05000293/2011-004-03

NCV Failure to Accurately

Assess Risk of Maintenance

on Standby Gas and Secondary

Containment

Closed

05000293/2011-001-00

LER Technical

Specification (TS) Required Shutdown -Reactor Building Closed Cooling Water (RBCCW)'B' Declared Inoperable

05000293/2011-002-00

LER Reactor Scram During a Planned Reactor Cool-Down with All Control Rods Fully Inserted

05000293/2011-003-00

LER Reactor Scram on Intermediate

Ranger Monitor High-High

Flux

05000293/2011-004-00

LER Technical

Specifications (TS) Required Shutdown Drywell to Torus DP Attachment

LIST OF DOCUMENTS

REVIEWED Section 1R01 Finat Safety Analysis Report, Section 2.4.4, Storm Flood Protection

CR-PNP-2010-3006, Coastal Storm Procedure does not direct a preemptive walkdown of the site Master/Local Control Center Procedure No. 2 (M/L CC2) Abnormal Conditions Alert, Revision 14 Procedure

2.1.37, Revision 28, Coastal Storm - Preparations and Actions Procedure

5.2.2, Revision 32, High Winds (Hurricane)

Procedure

2.1.42, Revision 10, Operation During Severe Weather Section 1R04

CR-PNP-201

1-3391 , Potential Seismic Interaction Hazard in the 'B' RHRyCS Quad

CR-PNP-201

1-3468, Step Ladder found in 'A' RHR Quad near lnstrumentation Rack

CR-PNP-2011-3770, Reactor Building Sump Pumps lying on the Floor Adjacent to the HPCI lnstrument Track Drawing M243, P&lD, HPCI System, Revision 53 Drawing M244, P&lD, HPCI System, Revision 31 Open Work Order Spread Sheet for System 23, HPCI Procedure

8.M.2-2.4.1, Revision 24, Core Spray Header Delta-P Procedure

2.2.19, Revision 104, Residual Heat Removal Procedure

8.C.43, Revision 10, Monthly System Valve Lineup Surveillance

Procedure

2.2.8, Revision 97, Standby AC Power System (Diesel Generators)

Procedure

2.2.22, Revision 72, Reactor Core lsolation Cooling System (RCIC)Procedure

2.2.21, Revision 79, High Pressure Coolant Injection System P&lD RCIC System M245, Revision E35 and M346, Revision E32 Third Quarter HPCI System Health Report Vendor Manual V-0257, HPCI Turbine, Revision 36 Vendor Manual V-0321, High Pressure Coolant Injection System, Revision 2 Section 1R05 Fire Hazards Analysis Final Safety Analysis Report, Chapter 10.8, Fire Protection System Fire Brigade 2011 Matrix for Operations' Participation in Fire Drills

CR-PNP-2011-3795, Power Panels in EDG building are not labeled correctly

CR-PNP-2011-3787, Diamond Plating installed on scaffolding potentially obstructing Overhead Fire Suppression

CR-PNP-2011-3794, Found Breaker in lighting panel tripped

CR-PNP-2011-3796, Severe kink in CO2 Hose Reel

CR-PNP-2011-3789, Procedure

5.5.2, Fire Plan Sheets are missing some Fire Protection Attributes

CR-PNP-2011-3790, Severe bend in CO2 Hose Reel

CR-PNP-2011-3793, 'A' Battery Room eyewash station is not secured Procedure

5.5.2, Revision 46, Special Fire Procedure Procedure

ENN-DC-189, Revision 1, Fire Drills Procedure

EN-TQ-125, Revision 0, Fire Brigade Drills Attachment Section 1R06

CR-PNP-2Q11-3497, Dam at entrance to Diesel Fire Pump room is separating from the wall

CR-PNP-2O11-3729, Potential Failure of Class ll Piping could affect'C'

SSW Pump Probabilistic Safety Assessment

IPE Update, Revision 1, Appendix E, Internal Flooding Analysis Final Safety Analysis Report, Revision 27, Sections 10 & 1 1 , Auxiliary System and Power Conversion System WO#

285591, Replace Dam in Diesel Fire Pump Room Regulatory Guide 1.29, Seismic Design Classification

TDBD-118, Revision E1, Design Basis Document for Seismic Design Updated Final SafetyAnalysis Report, Section 12.2.3.5, Seismic Loads Section 1R07 Final Safety Analysis Report, Revision 27, Section 10.7, Salt Service Water System and Section 10.5 Calculation

M-710, Revision 0, RBCCW Fouling Calculation

'A' RBCCW Thermal Performance Test for

RFO-17,

RFO-15, and

RFO-13 Procedure

8.5.3.14.1, RBCCW Thermal Performance Test Section 1Rl1 Combined Functional Drill 11-03, Hostile Action Based Scenario Emergency Action Levels for Security Threat Section 1R12

CR-PNP-2011-3049, Control Room Access Door Latch Failure and Functional Failure Determination

CR-PNP-2O11-3424,

CV-9068 A&B were found open on C903

CR-PNP-2g11-2538, Unable to establish Drywell to Suppression Chamber Differential Pressure

CR-PNP-2Q11-2993,Inaccurate supporting basis for Maintenance Rule Functional Failure Determination in

CR-PNP-201

-2538

CR-PNP-2011-3210, The Drywell to Torus Vacuum Breaker System Exceeded Maintenance Rule Unavailability

CR-PNP-2g11-3799, Periodic Update of Maintenance Rule Basis Documents has not been Completed

CR-PNP-201

1-3126, Conditions Associated with Door

DR-1 50 should have been Evaluated for Maintenance Rule Functional Failures

CR-PNP-2O11-3470, 'B'RBCCW HX reading not as expected

CR-PNP-2011-1330, Functional Failure Determinations, CA 1Q2,'B' RBCCW DP

CR-PNP-2011-3636, RBCCW HX dp evaluation unable to achieve 3500gpm

CR-PNP-2011-3663, 'A' and 'B' RBCCW HX failed weekly 2'2'32 Att'7

CR-PNP-201

1-3988, Functional Failure Determinations Result in RBCCW exceeding

a(1) criteria CR-pNP-2g11-3920,2.2.32

Att. 7 Fouling Evaluation resulting in backwashing

'A' and 'B'HX twice 10 cFR 50.65, Requirements for Monitoring at Nuclear Power Plants

EC 30600, Revision 2, Calculation

C15.0.2805, Affects of Control Room Ventilation on Vital Area Doors with Failed Latching Mechanisms

EN-DC-207, Revision 2, Maintenance Rule Period Assessment Expert Panel Meeting,

9130111 Agenda, SSW System 29 a(1) plan approval Attachment Functional Failu re Determ ination for

CR-PN P-2 01 1 -3424 Maintenance Rule Committee Meeting Minutes Dated

8116111 Maintenance Rule (aX1) Evaluation for CR-PNP-2011-3210

Maintenance Rule Periodic Assessment August 2009 P&lD HPCI System, M243, Sheet 1, Revision 53 P&lD HPCI System, M244, Sheet 1, Revision 1 Pilgrim Maintenance Rule Periodic Assessment Conducted June 20 - June 24,2011 Section 1Rl3

CR-PNP-201

1-3965, Line 342, Carrier Receiver is receiving

a Continuous Block Signal

CR-PNP-2011-3791, Qualitative Assessment of Risk for Standby Gas did not Result in an Increase of Risk One Color Level (to Yellow)Control Room Logs and Daily Risk Sheet for

813111 Equipment out of Service (EOOS) Quantitative Risk Assessment Tool Daily Risk Sheet for

7114111 Procedure

3.M.3-47, Revision 80, Attachment

5, Functional Test of Initiation Circuit Associated with'A' CSCS Pumps Procedure

3.M.3-47, Revision 80, Load Shed Relay Operational

/ Functional Test Procedure

1.5.22, Revision 14, Risk Assessment Process Procedure

EN-WM 104, Revision 4, On Line Risk Assessment

FSAR, Section 5.3, Secondary Containment Technical Specifications Section 3.7.C Basis Section for Secondary Containment Section 1R15

CR-PNP-2011-0334, RHR Loop 'A' Containment Spray Header Flow Transmitter Power Supply Ripple Voltage Out of Specification

CR-PNP-2011-3501, Compensatory Measure was required for

CR-PNP-2011-3344

and the operability evaluation was not completed in the specified timeframe

CR-PNP-2011-3424, HPCI Turbine Exhaust Line Drain Valves Open when they are Normally Closed

CR-PNP-2 01

1-1330, CA 1 03,

CR-PNP-201

-3424 Maintenance Rule Functional Failure Evaluation

CR-PNP-2011-3733, Failure to Include Seismic Input in channel-Control Blade Interference Guidance

CR-PNP-2011-4164, 'B' SBLC Accumulator has required charging every two weeks and associated operability evaluation

CR-PNP-2011-4200, SBLC tank high level alarm received in the control room and associated operability evaluation

CR-PNP-2011-4395, When restoring

'B' SBLC Accumulator, header pressure was reading 450lbs.EN-OP-104, Revision 5, Operability Determination Process FSAR, Section 6.4, High pressure Coolant lnjection System GE Hitachi 10 CFR Part2l Communication dated August 11,2011, Part 21 Reportable Condition Notification:

Failure to Include Seismic Input in Channel-Control Blade Interference Customer Guide GE Hitachi

Memorandum dated August 18,2011, Modification of Recommendations lssued in SC11-04 Procedure

8.E.10, Revision 45, LPCI System Instruments Calibration

Section 1Rl9

CR-PNP-2011-3007,Intermittent Fault causing Recirculation Flow Converter Failure Alarm

CR-PNP-2O11-3482, Discoloration on Pump Side Coupling on P-202D

CR-PNP-2011-3573, Air Line Feeding the

AO-N-98, Damper Broke during lnstallation

CR-PNP-2O11-3655, Incorrect Revision to System Drawing

CR-PNP-2011-2789, Procedure

8.C.4 does not test

AO-N-98 damper position

CR-PNP-2O11-4222, Review of Procedure

3.M.3-36.8

Revealed Missing Signatures

CR-PNP-2011-4228, Procedure

3.M.3-36.8

does not include a current check as a post installation step

CR-PNP-2O11-4263, WO# 290081-4 did not specify PWT current measurement as specified in

EN-WM-107

CR-PNP-2O11-4264, Procedure

3.M.3-71 needs to be revised

CR-PNP-2011-4127, Trip times on 2 of 3 poles for

ACB-103 were faster than 80-1 00MS acceptance criteria

CR-PNP-2011-4429, Post Maintenance Test Acceptance Criteria not Specified in Work Order for

ACB-103

CR-PNP-2O11-4431, Incorrect drawing referenced in Work Order for

ACB-103

EC 26526,

AO-N-98, Damper Replacement

EC 23892,

AO-N-98, Temporary Modification

EC 24796,

AO-N-98, Temporary Modification Change Notice Preferred MetalTechnologies Dynamics Operation and Pressure Drop Test Data Procedure

8.M.2-3.6.5, Revision 37, Recirculation Loop Instrumentation Neutron Monitoring Power Range Equipment Procedure

3.M.4-53, Revision 6, Check Valve Disassembly and Inspection

Procedure

8.C.4, Revision 24, Routine Running of Standby Gas Treatment System Procedure

8.M.2-2.3.1, Revision 31,

ADS-Pump Discharge

AC lnterlock Procedure

EN-WM-1 07, Revision 3, Post-Maintenance Testing Procedure

3.M.3.36.8, Revision 1, Temporary Power for +l- 24V DC Bus 'A' or'B'Procedure

3.M.3-71, Revision 2, lnspection and Maintenance of 345KV Disconnects, Insulators and Miscellaneous Switchyard Components Work Order (WO)#

00281250, Task 3, Perform Post Maintenance Testing for

FC-28-B WO#

00281250, Task 1,

FC-28-B Analog Output is Drifting and Spiking High WO#

52303161, Task 1 , Replace RBCCW Check Valve Internals WO#

5202681901 , Inspect Demister Drains

VGTF-2O18 WO#

5202681902, Post Maintenance Testing Operations

45-HO-48 WO#

5203478801, Inspect Demister Drain in

VGTF-2O1A WO#

5203478802, Post Maintenance Testing Operations

45-HO-44 WO#

0024642601,

EQ 26526 Replace Broken Damper WO#

0024642606, PWT after Damper Replacement (OPS)WO#0024642617, PWT after Damper Replacement (MECH)W O#

0022332501, Replace PS- 1 00 1 -938 Drain Valve W O#

0022332502, lsolate, Vent, Drain PS- 1 00 1 -938 WO#

0022332503, Post Work Testing of PS-1001-93B

WO#

290081, Tasks 3 and 4, D25, Post-Work Testing, Temp. Power to D25 WO#5231401301, Line 342, Switchyard insulator and Structure inspections

WO#

5231397701,

ACB-103 Breaker lnspection and Testing WO# 52313977Q3,

ACB-103 Breaker Inspection and Testing Post Work Test WO#

5231161401, Perform

ACB-103 Current Transformer Testing / Inspection

Section 1R22

CR-PNP-2011-3781, HPCI Drain Valves Opened forTwo Minutes Priorto Cold Start Surveillance has the Potential for Preconditioning

CR-PNP-2011-4106, Incorrect Surveillance Test Data Recorded in Procedure

8.5.3.14 Final Safety Analysis Report, Chapter 10.7, Salt Service Water System Procedure

8.5.3.1, Revision 59, RBCCW System Quarterly and Biennial Comprehensive OperabilitY

Procedure

CEP-IST-4, Revision 305, Standard of lnservice Testing Procedure

8.5.1 .1 , Revision 57, Core Spray System Operability - Pump Quarterly and Biennial Comprehensive Flow Rate Tests and Valve Tests Procedure

8.5.1.3, Revision 29, Core Spray Motor-Operated Valve Quarterly Operability Test'D'RBCCW Pump IST Vibration, Flow and Head Data Procedure

8.5.4.1-1, Revision 23, High Pressure Coolant Injection Simulated Automatic Actuation, Flow Rate and Cold Quickstart Test Procedure

8.5.3.14, Revision 32, Salt Service Water Flow Rate Operability Test Procedure

8.5.3.2.1, Revision 24, Salt Service Water Pump quarterly and Biennial (Comprehensive)

Operability and Valve Operability Tests NRC Inspection Manual Part 9900: Technical Guidance, Maintenance - Preconditioning of Structures, and Components before Determining Operability Control Room Logs for

712512011 WO#52290862, Task 1 Perform HPCI Cold Start Test Section 1EP2 RFQ# NP00121, Specifications for the Prompt Alert Siren Notification System for the Pilgrim Nuclear Power Station

EP-AD-417, Revision 4, Annual Siren Test Program

EP-AD-418, Revision 11, Monthly Testing of the Prompt Alert and Notification System (PANS)Ep-AD-419.

Revision 9, Annual Maintenance of the Prompt Alert and Notification System PANS Monthly Maintenance Forms, January 2010 - July 2011 PANS-related Condition Reports, January 2010 - July 2011 Pilgrim Nuclear Power Station Safety Evaluation for Emergency Action Levels (TAC No. ME0101)dated July 30, 2009 Section 1EP3

EP-PP-g1, Revision 36, Pilgrim Nuclear Power Station Emergency Plan, Section B: Station Emergency Organization Ep-PP-91, Revision 36, Pilgrim Nuclear Power Station Emergency Plan, Section E: Notification Methods and Procedures Ep-PP-91, Revision 36, Pilgrim Nuclear Power Station Emergency Plan, Section O: Emergency Response Training

EN-EP-801, Revision 2, Emergency Response Organization

EP-AD-410, Revision 3, Maintenance of the CANS Ep-AD-411, Revision 7, Testing of the Computerized Automatic Notification System (CANS)EN-PL-140, Revision 1, Emergency Response Organization Respiratory Protection Guidelines

PNPS Nuclear Training Manual, Revision 35

PLP-CHRP-EMER, Chemistry Technician Training Lesson Plan, Revision 3 pLp-CHRP-EMERRQL, Chemistry Technician

RP Duties during Declared Emergencies Requal, Revision 0 Attachment

PNPS ERO Roster (dated July 25, 2011)Test Sheets for PNPS Weekly Off-Hour Unannounced

ERO Notification Test, for all weeks between June 1 4,2011, and July 18,2011 Section 1EP4

EP-PP-Q1, Revision 36, Pilgrim Nuclear Power Station Emergency Plan

EN-LI-100, Revision 1 0, Process Applicability Determination

EN-EP-305, Revision 2, Emergency Planning 10CFR50.54(q)

Review Program EP-lP-100.1, Revision 8, Emergency Action Levels,

EP-AD-601, Revision 0, Emergency Action Level Technical Bases Document DIE Reviews for EP Procedures:

Emergency Plan lmplementing

Procedure

(EPIP) 260, Revision 1, EOF Operations

EPIP 261, Revision 2, TSC Operations

EPIP 300, Revision 8, Off-Site Rad Dose Assessment

EPIP 310, Revision 9, Off-Site Monitoring Team Activation and Response EPIP 400, Revision 14, Protective Action Recommendations

EPIP 520, Revision 7, Transition and Recovery DIE Reviews non-EP Procedures:

EN-EP-311, Emergency Response Data System (ERDS) Activation via Virtual Private Network (VPN), Revision 0-A

EN-EP-310, Emergency Response Organization Notification System, Revision 0

EN-EP-202, Equipment lmportant to Emergency Preparedness, Revision 0

EN-EP-8O1, Emergency Response Organization, Revision 0

EN-EP-308, Emergency Preparedness Critiques, Revision 1

EN-EP-306, Drills and Exercises, Revision 1

EN-TQ-110, Emergency Preparedness Training Program, Revision 0

EC 17120, ARINC Integration Platform, Including the SOCA Monitoring Center, Revision 0

EC-16895, Civil yard modifications

SOCA Enhancements at Pilgrim including Fences, VBS, MAC8X, and Underground Pathways, Revision 0 50.5a(q) screenings conducted between November 2010 and July 201 1 Procedure

5.3.14, Security lncidents, Revision 39 Procedure

2.4.143, Shut Down From Outside the Control Room, Revision 45 Procedure

5.2.1, Earthquakes, Revision 33 Section 1EPS Quality Assurance Audit Report

QA-07-201

0-PNP-1 (1 0CFR50.54(t)

Report)Quality Assurance Audit Report

QA-07-20 1 1 -PN P- 1 ( 1 0CFR50. 54(t) Report)Quality Assurance Surveillance Report

QS-201O-PNPS-020 (Evaluation of EP Interface between Taunton Emergency Management Director and PNPS)LO-PNPLO-2Q10-0070, Emergency Preparedness Exercise Readiness

LO-PNPLO -201

1-0020, Pre-NRC Inspection

NA 10-043, November 16,2Q10, NRC/FEMA EP Evaluated Exercise Report (10-05)NA 11-005, December 14,2010, Accountability Drill

NA 11-011, February g,2Q11,EP

Combined Functional Drill Report(1-01)

Quality Assurance Oversight Observation Checklists:

O2C-PNPS-2010-0059;

-0060; -0063; -0082; -0215 O2C-PNPS -201 1 -01 44: -01 45 Attachment

EP-related

Condition Reports

written between January 2010 and July 2011 Specific CRs reviewed:

CR-PNP-2O10-0521; -Q740; -1232; -1256; -1387; -1437; -1451

CR-PNP-2011-0383;

-0692; -0707; -0710; - 0838; -0862; -0970; -1188; -1462;-1489; -2440 Section 1EP6

CR-PNP-2011-4116, Alert EAL Classification during Emergency Preparedness Drill Needs to be Reviewed Combined Functional Drill 11-03, Hostile Action Based Scenario EP Combined Functional Drill Report (11-03), September

7,2011 Emergency Action Levels for Security Threat Performance Indicator Submittal Data Sheets Section 2RS08

CR-PNP-2010-0069

CR-PNP-2010-1343

CR-PNP-2010-2968

CR-PNP-2010-3025

CR-PNP-2010-2969

CR-PNP-2010-3256

CR-PNP-2010-2970

CR-PNP-2010-3607

CR-PNP-2010-0075

CR-PNP-2010-3690

CR-PNP-2010-1

105 CR-PNP-2010-3695

CR-PNP-2010-1214

CR-PNP-2010-3720

Procedure

EN-RW-102, Revision 8, Radioactive Shipping Procedure Procedure

EN-RW-104, Revision 8, Scaling Factors Procedure

EN-RW-105, Revision 1, Process Control Plan Procedure

EN-RP-108, Revision 10, Radiation Protection Postings Procedure

2.5.1.10, Revision 19, Transfer of Resin and Dewatering Liners Using Studvik Processing Facility THOR Dewatering System Procedure

2.5.1.11, Revision 9, Transfer of Sludge or Bead Resin and Dewatering

HIC Liners Using Studvik Processing Facility THOR Dewatering System Shippinq Packaqes: 1 1-05 Type A 1.28 Ci 11-08 Type A 4.06 Ci 11-09 Type A 16.7 Ci 10 CFR Part 61 Analvses Bead Resin Waste Stream Dry Activated Waste (DAW) Waste Stream Audits and Assessments:

RP Records / Dose Control 1l21l2O1O Shipping Exterior RAM Control 4112-1612010

LO10-069 Focused Assessment

8116-2012010

LO11.0137

>1FUhr @ 3 meters Radwaste Shipments

611412011

QA-14115-2009-PNP-01

RadiationProtection/Radwaste September

- October 02, 2009 Attachment Section 4OAl

EN-EP-201, Revision 12, Performance lndicators, DEP Pl data, October 2010 - June 2011 ERO Drill Participation Pl data, October 2010 - June 2011 ANS Reliability Pl data, October 2010 - June 2011 Licensee Event Reports MSPI Data Sheets from 3'd Quarter 2010 to 2nd Quarter 2011 for HPCI/RCIC

NEI 99-02, Revision 6, Regulatory Assessment Performance Indicator Guideline NRC Pl Approved Frequently Asked Questions NRC Performance lndicator Data Graphs NRC lnspection Reports MSPI Data Sheets from 3'd Quarter 2010 through 2no Quarter 2011 for RHR Control Room Logs RHR System Health Report Section 4OA2 Compensatory Actions and Disabled Annunciator Logs

CR-PNP-2011-4369, Disabled Annunciator Log Index in Error

CR-PNP-2O11-4367, Definitions of OperatorWorkAround and Operator Burdens Do Not Agree Between Site and Corporate

Procedures

CR-PNP-2011-4441, Discrepancy was identified with the Pl associated with the Outage Operator Workaround indicator for the month of August 201 1 Procedure

EN-FAP-OP-006, Revision 6, Operator Aggregate lmpact Index Performance Indicator Procedure

1.3.34.4, Revision 17, Compensatory Measures Pitgrim Operator Workarounds Aggregate lmpact Report Pilgrim Operator Compensatory Measures Log Pilgrim Open Operations Aggregate

Work Orders

Section 4OA3

LER 2011-001-00, Technical Specification Required Shutdown-

RBCCW'B' Declared Inoperable

LER 2011-002-00, Reactor Scram During a Planned Reactor Cool-Down with All Control Rods Fully Inserted

LER 2011-003-00, Reactor Scram on Intermediate Range Monitor High-High Flux

LER 2011-004-00, Technical Specification Required Shutdown-

Drywell to Torus DP

CR-PNP-2011-2538, Unable to Establish Drywell to Suppression Chamber Differential Pressure

CR-PNP-2011-3436, Shortfalls ldentified during the Conduct of the

712112011

Downpower IPTE Brief

CR-PNP-2011-4285, During development of Operability Evaluation for

CR-PNP-2011-4182, SSW motor currents were measured and are operating beyond full load capacity

CR-PNP-2011-4182, Multiple Motor Starts of RBCCW and SSW Pump Motors Power Maneuver Plan Approval Form Date/September

19,2011 Power Maneuver Plan Graph of Power Versus Date/Time for September

20,2011 Control Room Logs for

9122111 Emergent Risk Assessment for

9122111 Attachment

ADAMS ALARA ANS CA CR DRP DRS EAL EDG EP EPIP ERO FSAR HPCI HX rMc LER NCV NEI NRC PI PNPS QA RBCCW RCtC RFO RHR RPM RPS RWP SBGT SSC UFSAR WO A-11

LIST OF ACRONYMS

Agencywide

Documents

Access and Management

System as low as reasonably

achievable

alert and notification

system corrective

action condition

report Division of Reactor Projects Division of Reactor Safety emergency

action level emergency

diesel generator emergency

preparedness

emergency

plan implementing

procedure Emergency

Response Organization

final safety analysis report high pressure coolant injection heat exchanger inspection

manual chapter Licensee Event Report non-cited

violation Nuclear Energy lnstitute Nuclear Regulatory

Commission

performance

indicator Pilgrim Nuclear Power Station quality assurance reactor building closed cooling water reactor core isolation

cooling refueling

outage residual heat removal Radiation

Protection

Manager reactor protection

system radiation

work permit standby gas treatment structure, system or comPonent Updated Final Safety Analysis Report work order Attachment