IR 05000293/2006003

IR 05000293-06-003; 04/01-06/30/2006; Pilgrim Nuclear Power Station; Maintenance Risk Assessments and Emergent Work Control

ML062130615
Person / Time
Site:	Pilgrim
Issue date:	08/01/2006
From:	Racquel Powell NRC/RGN-I/DRP/PB5
To:	Balduzzi M Entergy Nuclear Operations
References
IR-06-003
Download: ML062130615 (31)

Text

August 1, 2006Mr. Michael A. BalduzziSite Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill RoadPlymouth, MA 02360-5508 SUBJECT:PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/2006003

Dear Mr. Balduzzi:

On June 30, 2006, the US Nuclear Regulatory Commission (NRC) completed an inspection atyour Pilgrim reactor facility. The enclosed integrated inspection report documents theinspection findings, which were discussed on June 30, 2006, with you and members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, reviewed your emergency preparedness program, observed activities, and interviewed personnel.This report documents one NRC-identified finding of very low safety significance. The findinginvolved a violation of NRC requirements which was classified at Severity Level IV inaccordance with the NRC's Enforcement Policy. However, because of the very low safetysignificance and because the issue has been entered into your corrective action program, the NRC is treating the issue as a non-cited violation (NCV), in accordance with Section VI.A.1 ofthe NRC's Enforcement Policy. If you contest the NCV in this report, you should provide aresponse with the basis for your denial, within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at Pilgrim. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosures 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 documentsystem (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/Raymond J. Powell, ChiefProjects Branch 5 Division of Reactor ProjectsDocket No. 50-293License No. DPR-35Enclosure: Inspection Report 50-293/06-03w/Attachment: Supplemental Information

SUMMARY OF FINDINGS

IR 05000293/200603; 04/01-06/30/2006; Pilgrim Nuclear Power Station; Maintenance RiskAssessments and Emergent Work Control.The report covered a 13-week period of inspection by resident inspectors, an announcedinspection by a regional specialist in health physics, and in-office reviews of emergency plan changes and grid reliability issues. One finding, which was a non-cited violation (NCV), wasidentified. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using IMC 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC managementreview. The NRC's program for overseeing the safe operation of nuclear power reactors isdescribed in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.A.Inspector Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Severity Level IV. The inspectors identified a Severity Level IV Non-Cited Violationassociated with the licensee's failure to perform an adequate safety evaluation per 10 CFR 50.59. Contrary to 10 CFR 50.59, a screening safety evaluation for handling of a 35 ton cask in the Reactor Building did not provide an adequate basis to demonstrate that the evaluation for use of a heavier cask did not change theevaluation methods approved by the NRC staff in 1985 for the control of heavy loadsper NUREG 0612 commitments, as described in the UFSAR and the Pilgrim licensingbasis. The licensee made significant enhancements to the original 50.59 safety evaluation and entered this issue into the corrective action program.The finding was determined to be more than minor because the inspectors could notreasonably determine that the methodology used to evaluate the use of a heavier caskdid not constitute a change that would have required NRC approval. The conditionsassociated with the finding (i.e., the potential drop of a loaded cask) were determined to be of very low safety significance because they did not result in the loss of operability of a safety system. Because the issue affected the NRC's ability to performits regulatory function, this finding was evaluated using the traditional enforcement process and was classified at Severity Level IV because the violation of 10 CFR 50.59 involved conditions evaluated as having very low safety significance by the SDP. This finding has a cross-cutting aspect in the area of human performance because Entergy did not fully evaluate the licensing basis to develop the 50.59 safety evaluation, andthereby failed to assure a design document was complete and accurate.

(Section 1R.13)

B.Licensee Identified Violations

None.

REPORT DETAILS

Summary of Plant StatusPilgrim Nuclear Power Station operated at 100 percent (%) core thermal power for the entirereport period, except for short periods of planned operation at reduced power for routine

===testing and maintenance.1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R04Equipment Alignment (71111.04).1Partial System Walkdowns

c. Inspection Scope

(4 samples)===

The inspectors completed a partial system review of the risk significant systems listedbelow to determine whether the systems were correctly aligned to perform theirdesigned safety function. The position of key valves, breakers, and control switches

===required for system operability were verified by field walkdown and/or review of themain control board indicators. To ascertain the required system configuration, theinspectors reviewed plant procedures, system drawings, the Updated Final SafetyAnalysis Report (UFSAR), and the Technical Specifications (TS). The references used for this review are listed in the attachment to this report. This inspection activity represented four samples.*High pressure core injection (HPCI) system during reactor core isolationcooling (RCIC) testing on April 11-12, 2006;*Residual heat removal (RHR) "A" and "B" trains, following maintenance andtesting on April 24, 2006;*RCIC system during HPCI testing on May 22-23, 2006; and *"B" Emergency Diesel Generator (EDG) during "A" EDG maintenance on June 14, 2006.

b. Findings

No findings of significance were identified.

2Enclosure1R05Fire Protection (71111.05).1 Quarterly Fire Protection Inspection

a. Inspection Scope

(9 samples)===

The inspectors toured selected areas of the plant to observe conditions related to:

(1) transient combustibles and ignition sources;

(2) fire detection systems;

(3) manualfirefighting equipment and capability; and

(4) passive fire protection features. The inspectors verified adequate material condition of active and passive fire protection systems features and their operational line up and readiness. Theinspectors also reviewed the applicable fire hazard analysis fire zone data sheets.

The references used for this review are listed in the attachment to this report. This inspection activity represented nine samples.*Fire Zone 1.16, Reactor Building Open Area at El. 91, North Half;*Fire Zone 1.20, Refueling Floor;

• Fire Zone 1.18, Contaminated Equipment and Skimmer;

• Fire Zone 1.17, Clothing Change Area;

• Fire Zone 1.15, Standby Liquid Control;

• Fire Zone 1.10B, B RHR and HPCI Pipe Room;*Fire Zone 2.1, B Switchgear and Load Center Room;

• Fire Zone 4.1, B Train Diesel Generator Room; and

• Fire Zone 4.2, B Train Diesel Day Tank Room.

b. Findings

No findings of significance were identified..2Annual Fire Drill Observation n.Inspection Scope (1 sample)The inspectors observed a training fire drill conducted on April 28, 2006 per procedure1.4.23, "Fire Brigade Training Drill." The unannounced drill involved a simulated fire inthe onsite two-story Butler Building, which contained a simulated radiologically controlled area. The unannounced drill involved the combined response of the onsitefire brigade (FB) and the Plymouth Fire Department (PFD). The inspectors observed fire personnel performance, and confirmed that the licensee's fire fighting pre-plan strategies per procedure 5.5.2, "Special Fire Fighting Procedure," were utilized, thepre-planned dr ill scenario was followed, and the drill objectives were met. Theinspectors verified the joint use of the Incident Command System by the FB and PFD.

The inspectors confirmed that proper security and radiological controls were applied; proper protective clothing and breathing apparatus were donned; sufficient fire fightingequipment 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 confirmed the drill critique identified areas to 3Enclosureenhance fire brigade performance. The inspectors verified that the licensee identifiedappropriate corrective actions for identified deficiencies and entered the issues into the corrective action program.

This activity represented one inspection sample. o.FindingsNo findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11).1Licensed Operator Simulator Training

a. Inspection Scope

(1 sample)The inspectors observed the performance of an operator crew during a simulatortraining session on May 11, 2006. The training was conducted as part of licensed operator just-in-time (JIT) training for the planned movement of a 35 ton shipping cask inside the reactor building. Licensee evaluations in UFSAR Section 10.3.6 concluded that the postulated drop of the shipping cask could constitute a severe plant event thatwould result in an operational transient and impact plant safety systems. Theinspectors verified the JIT training scenario developed was adequate to ensure the crew's ability to safely shutdown the plant. The inspectors evaluated whether the crew met the training scenario objectives, performed the critical ta sks, and properly usedabnormal operating procedures and emergency operating procedures. The inspectors verified that the post-scenario critique discussed items for improvement with the crew to enhance performance. The references used for this review are listed in the attachment to this report. This inspection activity represented one sample.

b. Findings

No findings of significance were identified..2Licensed Operator Simulator Exams

a. Inspection Scope

(1 sample)The inspectors observed an evaluated licensed operator simulator training exercise onMay 22, 2006. The training was performed using scenario SES-00-00-152 and involved both operational transients and design basis events. The inspectors evaluated both the crew's performance and evaluators' assessments. Specifically, the inspectors evaluated whether the crew met the scenario objectives, accomplished the critical tasks, demonstrated proper use of abnormal and emergency operatingprocedures, demonstrated proper command and control, communicated effectively, and implemented the emergency plan in-terms of event classification and notification. The inspectors reviewed the post-scenario critique and confirmed lessons learned and items for improvement were discussed with the crew to enhance future performance.

b. Findings

No findings of significance were identified.

1R12 Maintenance Rule (71111.12)

a. Inspection Scope

(3 samples)The inspectors reviewed follow-up actions for issues relating to the selected systemsand reviewed the performance history of the systems to assess the effectiveness ofEntergy's maintenance activities. The inspectors reviewed Entergy's problem identification and resolution actions for these issues in accordance with NRC procedures and the requirements of 10 CFR 50.65(a)(1) and (a)(2), "Requirements for Monitoring the Effectiveness of Maintenance." In addition, the inspectors reviewed system classification, performance criteria and goals, system health reports, andcorrective actions that were taken or planned to verify whether the actions were reasonable and appropriate. These inspection activities represented three samples: *Proper classification of equipment issues for System 50 - PrimaryContainment, including the Operational Decision Management Issue (ODMI)for drywell leakage (Condition Report (CR) 200503299). The inspectors reviewed Entergy's basis for placing the system in maintenance rule (a)(2)status.*Proper classification of equipment issues for System 54 - Reactor PressureVessel, including the core shroud tie down bolts (CR 200601849). *The inspectors reviewed Entergy's basis for placing the system in maintenancerule (a)(2) status.*Proper classification of equipment issues for System 66 - Process RadiationMonitors C19A/B. The inspectors reviewed Entergy's basis for placing the system in maintenance rule (a)(2) status.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

(5 samples)The inspectors evaluated on-line risk management for planned and emergent work. The inspectors reviewed maintenance risk evaluations, work schedules, recent corrective actions, and control room logs to verify that other concurrent planned and emergent maintenance or surveillance activities did not adversely affect the plant riskalready incurred with the out-of-service components. The inspectors evaluated whether Entergy took the necessary steps to control work activities, took actions to minimize the probability of initiating events, and maintained the functional capability ofmitigating systems. The inspectors assessed Pilgrim's risk management actions 5Enclosureduring plant walkdowns. The inspectors also discussed risk management activitieswith maintenance, engineering and operations personnel as applicable. References used for the inspection are identified in the attachment to this report. The inspectioncovered the following five samples: *The elevated (Yellow) risk associated with the RHR logic system functional teston April 17, 2006;*The elevated (Yellow) risk condition on May 11, 2006, associated with logicsystem functional test of the "A" EDG and "A" RHR system;*The elevated (Orange) risk associated with the inspection of salt service waterpump P208E per maintenance request (MR) 06107824 on June 1, 2006;*The elevated (Yellow) risk condition the week of June 11 for plannedmaintenance activities on the A emergency diesel generator; and*The risk associated with the control of heavy loads while handling a CNS 3-55waste shipment cask in the Reactor Building as part of the spent fuel pool cleanup activities: MR 05118433, "Clean Spent Fuel Pool, Ship Irradiated Hardware;" and, engineering request (ER) 05120679, "Provide NUREG 0612Heavy Loads Evaluation / Safe Load Path for the CNS 3-55 Shipping Cask and the Crusher Shearer Tool and Stand to be used in the Fuel Pool Cleanup Project."

b. Findings

Introduction:

The inspectors identified a Severity Level IV Non-Cited Violation associated with thelicensee's failure to perform an adequate safety evaluation (SE) as required by 10 CFR 50.59 for changes made to the facility.Description:The Pilgrim UFSAR describes the methods for controlling heavy loads and theevaluations used to determine the consequences of a dropped cask in the spent fuel pool. The methods were described in License Amendments 20, 24 and 29, which were incorporated in UFSAR Section 10.3.6. In response to NRC requests foradditional information in 1980, the licensee described additional evaluations and methods to meet the NUREG 0612 criteria to control heavy loads and mitigate theconsequences of a cask drop event. The evaluations were described in letters to NRC dated June 25, 1981; October 8, 1981; and July 13, 1983. The NRC staff approvedthe licensee's methods and evaluations in an SE dated March 6, 1985. The 1981 and1983 evaluations became part of the licensing basis established in UFSAR Sections 10.3.6 and 12.2.3.7 as the methods accepted by the NRC staff for the control of heavyloads. The loads evaluated for the NUREG 0612 licensing basis included a shippingcask with a loaded weight of 26 tons.Entergy planned to use a CNS 3-55 shipping cask to transport radioactive waste fromPilgrim Station during a spent fuel pool clean-up project in 2006. The CNS 3-55 cask 6Enclosurehas a maximum loaded weight of 35 tons. The licensee completed ER 05120679because "...shipping casks previously evaluated for licensing basis NUREG 0612compliance addressed a spent fuel shipping cask with a loaded weight of 26 tons.

Since the proposed CNS 3-55 cask is heavier than the approved casks, it must beevaluated for NUREG 0612 compliance." ER 05120679 further defined requirementsto assure that equipment and procedures used for the spent fuel pool cleanup were in accordance with the NUREG 0612 licensing commitments regarding safe load paths,lifting devices and load drop consequences.Entergy performed a 10 CFR 50.59 screening review for ER 05120679, dated March24, 2006. Entergy concluded that a complete 50.59 evaluation was not requiredbecause the proposed activity screened out based on a determination that ER05120679 did not change a "method of evaluation" described in the UFSAR or used in establishing the licensing basis. The inspector identified several issues which indicated ER 05120679 had made changes to the evaluation methods used in the licensing basis approved by the NRC staff in 1985, and questioned whether thelicensee needed prior NRC review per 10 CFR 50.59(c)(2)(viii).NRC concerns involved the movement of the cask in and out of the spent fuel pooland the movement of the cask across the operating floor. To evaluate the move of theCNS 3-55 cask across the refueling floor, ER 05120679 used energy balance methods to conclude that handling the 35 ton cask at a height of four inches above thefloor was equivalent to the licensing basis of handling the 26 ton cask at six inches.

ER 05120679 prescribed the use of wood cribbing to maintain the four inch distance when lifting the cask above interferences around the periphery of the spent fuel pool.

ER 05120679 relied upon a 1993 analysis to evaluate spent fuel pool integrity following a cask drop. The 1993 methodology had not received previous NRC reviewand approval as part of the NUREG 0612 evaluations. Further, the 1993 analysiscredited an impact limiter (Hexcel energy absorbing pad) in evaluating the cask drop consequences. The use of an energy absorber pad had not been reviewed by the NRC as part of the NUREG 0612 evaluations.The inspectors' concerns were reviewed with licensee staff in meetings on March 24,April 6, April 7, May 5 and May 10, 2006. In response, the licensee first added detailsto the bases for the 10 CFR 50.59 screening, and then further researched the complete licensing basis for the control of heavy loads which was incorporated into a full 50.59 safety evaluation. The safety evaluation was issued in ER 05120679, SE 3402 Revision 1, dated May 10, 2006. Following a review on May 10, the inspectors concluded the safety evaluation dated May 10, 2006 fully described the licensing basis, showed the relevance of the 1993 analysis without relying on it for the NUREG0612 commitments, and showed that a 35 ton cask and the Hexcel pad had been described in the licensing basis for the control of heavy loads. The licensee entered this issue into the corrective action program as CR 200602460.

7EnclosureAnalysis:A performance deficiency was identified in that Entergy had not developed an adequatebasis to support the 10 CFR 50.59 screening safety evaluation dated March 24, 2006.

The March 24 SE for ER 05120679 was inadequate because it did not fully describe the complete licensing basis for the control of heavy loads, and it used evaluation methods different than those approved by the NRC staff in 1985 for the control ofheavy loads per NUREG 0612. The finding was determined to be more than minorbecause the inspectors could not reasonably determine that the methodology used toevaluate the use of a heavier cask did not constitute a change that would have required

NRC approval. The conditions associated with the finding (i.e., the potential drop of aloaded cask) affected the objective of the Mitigating Systems cornerstone to ensure the availability of systems to respond to events. The conditions were assessed using theSDP and determined to be of very low safety significance because they did not result in the loss of operability of a safety system. Because the issue affected the NRC's abilityto perform its regulatory function, this finding was evaluated using the traditional enforcement process and was classified at Severity Level IV because the violation of 10 CFR 50.59 involved conditions evaluated as having very low safety significance by the SDP.This finding has a cross-cutting aspect in the area of human performance becauseEntergy did not fully evaluate the licensing basis to develop the 10 CFR 50.59 SE, andthereby failed to assure a design document was complete and accurate.Enforcement:10 CFR 50.59(a)(1) defines changes to the facility as described in the UFSAR toinclude changes to evaluations that demonstrate that intended functions will beaccomplished. 10 CFR 50.59(c)(1) states a licensee may make changes to the facilityand procedures as described in the UFSAR without obtaining a license amendment pursuant to 10 CFR 50.90 only if the change does not meet any of the criteria in paragraph (c)(2). 10 CFR 50.59(c)(2)(viii) states a licensee shall obtain a licenseamendment pursuant to 10 CFR 50.90 prior to implementing a proposed change if the change results in a departure from a method of evaluation described in the UFSAR used in establishing the design bases. 10 CFR 50.59(a)(2)(ii) defines departures from a method of evaluation as changing a method described in the UFSAR to another method, unless that method has been approved by NRC for the intended application. 10 CFR 50.59(d)(1) requires a written evaluation which provides the bases for the determination that the change does not require a license amendment. Contrary to the above, in a 50.59 screening evaluation for ER 05120679, datedMarch 24, 2006, Entergy failed to provide an adequate basis for the determination thatthe handling of heavy loads for the spent fuel pool cleanup project did not result in a change in the UFSAR method of evaluation per NUREG 0612 commitments asdescribed in UFSAR Sections 10.3.6 and 12.2.3.7. Because the violation is classifiedat Severity Level IV and has been entered into Entergy's corrective action program (CR 200602460), this violation is being treated as a Non-Cited Violation (NCV),

8Enclosureconsistent with Section VI.A of the NRC Enforcement Policy. NCV 0500293/2006003-001: Failure to perform an adequate 50.59 evaluation for the control of heavy loads.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

(4 samples)The inspectors reviewed selected operability determinations to assess the adequacy ofthe evaluations, the use and control of compensatory measures, compliance with the Technical Specifications, and the risk significance of the issues. The inspectors used the Technical Specifications, UFSAR, associated Design Basis Documents, Procedure ENN-OP-104 "Operability Determinations," and the additional references listed in theattachment to this report for Section 1R15. This review covered four inspection samples:*CR 200601849, General Electric Safety Communication 06-07, Core ShroudRepair Tie Rod Upper Support Cracking;*CR 200602122, Station blackout diesel generator jacket water expansion tankcontains small amount of oil;*CR 200602271, "A" EDG gear box back lash out-of-tolerance; and

• CR 200602222, "A" EDG inner and outer slip ring tolerance out-of-tolerance.The inspectors verified Entergy was identifying problems with operability determinationsat an appropriate threshold and entering them into the corrective action program.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

(7 samples)The inspectors reviewed post-maintenance test activities on risk significant systems toverify that the effect of the test on the plant had been evaluated adequately, the testwas properly performed in accordance with procedures, the test data met the required acceptance criteria, and the test activity was adequate to verify system operability andfunctional capability following maintenance. The inspectors confirm ed that systemswere properly restored following testing and that discrepancies were appropriately documented in the corrective action process. The inspection activity representedseven samples:*Replacement of RHR system motor operated valve fuses per MR Nos.:

05119912, 05119913, 05119914, 05119829, 05119830, 05119831, 05119832, 05119834, 05119835, 05119836, 05119851, 05119853, and 05120313; 9Enclosure*HPCI testing per licensee procedure 8.5.4.1 following high flow and temperaturetesting;*Seismic monitor calibration and functional testing following repairs per MR 05117438;*Replacement of "A" EDG governor droop relay per MR 06103224;

• Replacement of "A" EDG M2 starting air pressure regulator (PCV-4592) andsolenoid (SV-4586A) per MR 06102280;*Emergent work for Alarm 3L-D1 (Voltage/Frequency Abnormal) per MR 06104965; and*Testing of the "A" EDG following two, four, and six year preventive maintenanceactivities performed in accordance with licensee procedures 3.M.3-61.5, 3.M.3-61.9, and 3.M.3-61.10, respectively. MRs 04117873, 06108858, 06104077, 04107062, P9901178, 02114113, 05104966, 06108840, 04109587, 06107595, 06103632.The inspectors verified Entergy was identifying post-maintenance testing problems atan appropriate threshold and entering them into the corrective action program.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope

(8 samples)The inspectors observed and/or reviewed surveillance testing results to determinewhether the test acceptance criteria was consistent with Technical Specifications (TS)and related Performance Indicators (PI), that the test was performed in accordance with the written procedure, the test data was complete and met procedural requirements, and the components were capable of performing their intended safety functions. The inspection activity represented eight samples:*2.1.15, RCS Leakage Rate Measurements for April - May 2006;*8.5.2.2.1, LPCI System Loop "A" Operability - Pump Quarterly and Biennial(Comprehensive) Flow Rate and Valve Tests, 4/21/06;*8.5.2.3, LPCI and Containment Cooling Motor-Operated Valve Operability Test, 4/21/06;*8.M.2-2.10.2-16, LPCI Break Detection Logic Functional Tests Injection ValvesInterlock Test - Division "A";*8.M.2-2.10.2-17, LPCI Break Detection Logic Functional Tests Injection ValvesInterlock Test - Division "B";*8.5.4.1, HPCI System Pump and Valve Quarterly Test (IST), 5/23/06;

• 8.9.1, Emergency Diesel Generator and Associated Emergency BusSurveillance; and*8.M.2-2.10.8.1, Diesel Generator "A" Initiation by RHR Logic.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

(71111.23)

a. Inspection Scope

(1 sample) The inspectors reviewed Temporary Alteration 06-1-08 to verify that the licensing basesand performance capability of the associated risk significant system had not beendegraded through the modification. The references used for this review are listed in the attachment to this report. This inspection activity represented one sample.Temporary Alteration 06-1-08 installed a temporary 24 vdc power system for theneutron monitoring and process radiation monitoring instrumentation while replacing the existing 24 volt batteries. The licensee provided an analysis as part of the technical justification for TA 06-1-08. The inspectors discussed the temporary alteration with licensee personnel and observed work activities in progress. The inspectors reviewed the controls used by the licensee to assure the 24 vdc system remained operable. The inspectors reviewed the changes to applicable plant drawings and confirmed the modifications were installed per TA 06-1-08.

b. Findings

No findings of significance were identified.

===Cornerstone: Emergency Preparedness1EP4Emergency Action Level and Emergency Plan Changes (71114.04)

a. Inspection Scope

(1 sample)An in-office inspection to review recent changes to the Pilgrim Nuclear Power StationEmergency Plan (revision 32) was conducted on June 22 - 23, 2006. These changes===

were made in accordance with 10 CFR 50.54(q). The licensee had determined that the changes did not decrease the effectiveness of the Plan and concluded that the Plancontinued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR 50. During this inspection, the inspectors conducted a sampling review of the changes that could potentially result in a decrease in effectiveness. This review did not constitute an approval of the changes and, as such, the changes are subject to future NRC inspection. The inspection was conducted in accordance with NRC InspectionProcedure 71114, Attachment 4, and the applicable requirements in 10 CFR 50.54(q)were used as reference criteria.

b. Findings

No findings of significance were identified.1EP6Drill Evaluation (71114.06).1Event Classification During Operator Simulator Training

b. Inspection Scope

(1 sample) The inspectors observed an evaluated licensed operator simulator training exercise onMay 22, 2006, and evaluated the crew's ability to implement the emergency plan. Specifically, the inspectors confirmed the crew properly classified the event, activated the notificati on system, and appropriately completed and transmi tted the eventnotification forms in a timely manner.

c. Findings

No findings of significance were identified..2Combined Functional Drill

a. Inspection Scope

(1 sample) The inspectors reviewed the combined functional dr ill scenario (06-01) conducted onJune 1, 2006, and observed portions of the drill at the technical support center and theemergency operation facility. The inspection focused on the ability of Entergypersonnel to properly conduct classification, notification, and protective action recommendation activities, and on the evaluators' ability to identify observedweaknesses and/or deficiencies within these areas. The inspectors attended the player post-drill critiques to compare NRC identified deficiencies against the licensee'sidentified findings to determine whether Entergy was properly identifying weaknesses in these areas. The inspectors reviewed licensee actions to address issues in the corrective action program. The references used in this review included the Controller Manual Combined Functional Drill (06-01) dated J une 1, 2006.

b. Findings

No findings of significance were identified.

12Enclosure2.RADIATION SAFETY Cornerstone: Public Radiation Safety2PS1Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems(7112201)

a. Inspection Scope

(10 samples)The inspectors reviewed the most current Pilgrim Nuclear Power Station RadiologicalEffluent and Waste Disposal Report to verify that the program was implemented as described in the Radiological Effluent Technical Specification/Offsite Dose Calculation Manual (RETS/ODCM). The inspectors reviewed the report for significant changes to the ODCM and radioactive waste system design and operation to determine whetherthe changes to the ODCM were made in accordance with Regulatory Guide 1.109 and NUREG-0133 and were technically justified and documented, and to determinewhether the modifications made to radioactive waste system design and operationchanged the dose consequence to the public. The inspectors also verified that technical and design change reviews, such as 10 CFR 50.59 reviews, were performed as required and determined whether radioactive liquid and gaseous effluent radiation monitor setpoint calculation methodology changed since completion of the modifications. The inspectors also reviewed the report to assure that any anomalous information was effectively reported and explained. The inspectors reviewed the RETS/ODCM to identify the effluent radiation monitoring systems and associated flowmeasurement devices; reviewed effluent radiological occurrence performance indicator incidents for onsite follow-up; and reviewed licensee self assessments, audits, and licensee event reports that involved unplanned releases of radioactive material. The inspectors noted there had been no changes made by the licensee to the ODCM or to the liquid or gaseous radioactive waste system design or operation since the lastinspection in 2004. The inspectors walked down the major components of the gaseous and liquid releasesystems (e.g., radiation and flow monitors, demineralizers and filters, tanks, andvessels) to observe ongoing activities, current system configuration with respect to thedescription in the UFSAR, and equipment material condition. The inspectors reviewed the liquid discharge permit used since the previous inspection,including the projected doses to members of the public. The inspectors also observed the routine sample collection and analysis for the continuous release of radioactive gaseous effluent to verify that appropriate treatment equipment was effectively used and that the radioactive gaseous effluent was processed and released in accordance with RETS/ODCM requirements. The inspectors reviewed the release records to confirm that adequate controls were in place to prevent an unmonitored or unanticipated release of radioactive material to the environment.

13EnclosureThe inspectors reviewed a selection of monthly, quarterly, and annual dose calculationsto ensure that the licensee had properly calculated the offsite dose from radiological effluent releases and to determine if any annual Technical Specification/ODCM (i.e.,

Appendix I to 10 CFR Part 50) values were exceeded and, if appropriate, a PI report was issued.The inspectors reviewed air cleani ng system surveillance test results and licenseespecific methodology to ensure that the system was operating within the licensee'sacceptance criteria. The inspectors also reviewed surveillance test results and the methodology the licensee uses to determine the stack and vent flow rates and evaluated whether the flow rates are consistent with RETS/ODCM or UFSAR values. The inspectors reviewed records of instrument calibrations performed since the lastinspection for each point of discharge effluent radiation monitor and flow measurement device, and reviewed any completed system modifications and the current effluentradiation monitor alarm setpoint value for agreement with RETS/ODCM requirements.

The inspectors also reviewed calibration records for radiation measurement (i.e., counting room) instrumentation associated with effluent monitoring and release activities and reviewed quality control records for the radiation measurement instruments. The inspectors reviewed the results of the interlaboratory comparison program to verifythe quality of radioactive effluent sample analyses performed by the licensee; reviewed the licensee's quality control evaluation of the interlaboratory comparison test and associated corrective actions for any deficiencies identified; and reviewed the results from the licensee's QA audits to verify that the licensee met the requirements of the RETS/ODCM.The inspectors reviewed the licensee's Licensee Event Reports, Special Reports,audits, and self assessments related to the RETS/ODCM program performed since the last inspection. The inspectors confirmed that identified problems were entered into the corrective action program for resolution. The inspectors also reviewed corrective action reports related to environmental sampling, sample analysis, or meteorological monitoring instrumentation.

b. Findings

One unresolved item was identified related to the licensee's particulate samplingprocess for the reactor building vent and main stack.

Description:

TS 5.5.4.c requires the licensee to monitor, sample and analyzeradioactive effluents in accordance with the methodology and parameters in the ODCM. The ODCM, in section 7.2.2, Main Stack Gas Monitoring System, and section 7.2.3, Reactor Building Exhaust Vent Monitoring System, specifies that samples are drawn through an isokinetic probe which is located to assure representative sampling 14EnclosureThe inspectors requested information relative to the licensee's basis for the sampleflow rate range of 1.6 to 1.8 cfm, as specified in procedure PNPS 7.3.37 for the reactor building Accordingly, thismatter is considered unresolved pending completion of the licensee's analysis anddetermination of the consequence of this condition. URI 050000293/2006003-02:Anisokinetic sampling of reactor building vent and main stack gaseous effluents.

OTHER ACTIVITIES

[OA]4OA1Performance Indicator Verification (71151).1Reactor Safety Cornerstones

a. Inspection Scope

(3 samples)The inspectors reviewed PI data to confirm the accuracy and completeness of thereported data. The review was accomplished by comparing reported PI data to confirmatory plant records and data available in plant logs, the chemistry data base (WinCDMS), maintenance rule records, Licensee Event Reports, condition reports and NRC inspection reports. The inspection activity represents three samples.*Mitigating System Cornerstone, Safety System Functional Failures from thethird quarter of 2004 through first quarter of 2006;*Barrier Integrity Cornerstone, Reactor Coolant System Specific Activity from thethird quarter of 2004 through the first quarter 2006; and*Barrier Integrity Cornerstone, Reactor Coolant System Unidentified Leakagefrom the third quarter of 2004 through the first quarter 2006.

b. Findings

No findings of significance were identified.

15Enclosure.2Public Radiation Safety Cornerstone

a. Inspection Scope

(1 sample)The inspectors sampled licensee data for the RETS/ODCM Radiological EffluentOccurrences PI. PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Rev. 2, were used to verify the accuracy of the PI data reported.The inspectors reviewed the Radiological Control Effluent Release Occurrences PIresults for the Public Radiation Safety Cornerstone. For the assessment period, the inspectors reviewed selected out of service effluent radiation monitor and compensatory sampling data, any abnormal release results as reported in the 2004 and 2005 Annual Effluent Reports, procedural guidance for reporting PI information, and selected condition reports related to RETS/ODCM issues. In addition, the inspectors reviewed cumulative and projected doses to the public for the period October 2004 through May 2006. Documents reviewed are listed in sections 2PS1 and

4OA1 of the

report attachment.

b. Findings

No findings of significance were identified.4OA2Identification and Resolution of Problems (71152)1.Routine Review of Corrective Action Program Issues

a. Inspection Scope

As required by Inspection Procedure 71152, "Identification and Resolution ofProblems," 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 of significance were identified. 2.Corrective Action Program Semi-annual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, "Identification and Resolution ofProblems," the inspectors performed the semi-annual trend review to identify trends, 16Enclosureeither Entergy or NRC identified, that might indicate the existence of a more significantsafety issue. Included within the scope of this review were condition reports from October 2005 through June 2006, the 4 th quarter 2005 corrective action trend reports,and the daily plant status report listings of operations equipment problems, operabilityevaluations, and temporary alterations.

b. Findings and Observations

No findings of significance were identified. Several trends were identified, but nonethat suggested the presence of a more significant safety issue. The majority of the trends identified by the inspectors had been recognized by Entergy and captured in adverse trend CRs, including an emerging adverse trend in instrument air systemperformance (CR 2005-4706) which is being evaluated by Entergy. The inspectors identified that adverse trends noted by Entergy in the areas of communication equipment and meteorological tower performance appear to be further degrading.

Adverse trends not captured by the current licensee trend report were noted regarding augmented off-gas system spikes and/or pre-treat HI RAD alarms, expiredchemicals/reagents in the chemistry lab, and water in the station blackout fuel oil storage tanks. The licensee noted these items for further consideration.

.3 Annual Sample Review - Operator Workarounds

a. Inspection Scope

(1 sample)The inspectors reviewed the cumulative effect of operator workarounds on thereliability, availability, and potential mis-operation of systems with particular focus onissues that had the potential to affect the ability of operators to respond to planttransients and events. The inspectors reviewed the Operator Compensatory MeasureLog, the Operator Aggregate Impact Index for April 2006, and Operations PerformanceIndicators, as well as the related operator workarounds, operator burdens, control room deficiencies, system lineup deviations, protective and caution tagouts, and disabled orilluminated control room alarms. For selected issues, the inspectors discussed the issues with responsible operations personnel to ensure they were appropriately categorized, prioritized and tracked for resolution.

b. Findings and Observations

No findings of significance were identified. The inspectors found that Entergy ensuredthat appropriate attention was placed on conditions that could impact operator actions,including conditions that would require compensatory actions (workarounds and burdens), control room deficiencies and alarms, and components tagged out-of-service or with caution tags, through periodic management review of performance indicators. Appropriate actions were taken to ensure that operators were aware of the issues, and corrective actions were scheduled for completion commensurate with each item's significance.

17Enclosure4OA3Event Follow-up (71153)Licensee Event Report Review and Closeout (1 sample) a.(Closed) LER 05000293/2006-001-00, Manual Scram due to High Offgas RecombinerTemperature Resulting from Inadequate Preventive Maintenance of recombinerPreheater Pressure Control Valve Controller. The inspectors reviewed Entergy'sactions associated with Licensee Event Report (LER) 50-293/2006-001. Entergy's actions were addressed in the corrective action program as CR 20060977. The event was also described in NRC report 2006-002, which documented a Green NCV (NCV05000293/2006002-001). The LER provided an accurate description of the event and follow-up actions, taken or planned, were appropriate to address the event. This LER is closed.4OA5Other.1Implementation of Temporary Instruction (TI) 2515/165 - Operational Readiness of Offsite Power and Impact on Plant Risk

a. Inspection Scope

The objective of TI 2515/165, "Operational Readiness of Offsite Power and Impact onPlant Risk," was to gather information to support the assessment of nuclear power plant operational readiness of offsite power systems and impact on plant risk. The inspectorsevaluated licensee procedures against the specific offsite power, risk assessment and system grid reliability requirements of TI 2515/165. They also discussed the attributeswith licensee personnel. The information gathered while completing this TI was forwarded to the Office ofNuclear Reactor Regulation for further review and evaluation on April 3, 2006.

b. Findings

No findings of significance were identified..2Strike Contingency Planning (92709)

a. Inspection Scope

(1 sample)Entergy developed a staffing contingency plan to continue Pilgrim Station securityoperations should union personnel engage in a job action. Using the guidance of Inspection Procedure 92709, the inspectors reviewed licensee plans to address a potential job action. The inspection included an evaluation of the strike contingency plan content and the actions needed to implement the plan; and, a review to determine iffacility security would be maintained as required with a sufficient number of qualifiedpersonnel.

NRC review of this area continued at the end of the inspection.

b. Findings

No findings of significance were identified. 4OA6Meetings, Including ExitExit Meeting SummaryOn June 30, 2006, the inspectors presented the inspection results to members ofEntergy management led by Mr. Michael Balduzzi. The inspectors confirmed that there was no information that Entergy considered proprietary included in this report.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

S. BethayDirector, Nuclear Assessment

K. Bronson General Manager Plant Operations

G. DykemanDesign Engineering

B. FordLicensing Manager

B. GrievesQuality Assurance Manager

P. LeavittChemistry

D. LandecheSpecial Projects Manager

W. LoboLicensing Specialist

J. McClellanQuality Specialist-Quality Assessment

B. McDonaldRadiation Protection Specialist (Support)

P. McNultyRadiation Protection Manager

D. Noyes Assistant Operations Manager

E. OlsonOperations Manager

C. PittsDesign Engineer

M. SantiagoTraining Supervisor

K. SejkoraEffluent Engineer

D. SeligPrograms and Components Supervisor

J. TaorminaWork Control Supervisor

T. TraskSystem Engineering Manager

NRC personnel

W. Raymond, Senior Resident Inspector

C. Welch, Resident Inspector

LIST OF ITEMS

Opened

05000293/2006003-02URIAnisokinetic sampling of reactor building vent and mainstack gaseous effluentsClosed05000293/2006-001-00LERManual Scram due to High Offgas RecombinerTemperature Resulting from Inadequate Preventive

Maintenance of Recombiner Preheater Pressure Control

Valve Controller.

A-2AttachmentOpen and

Closed

05000293/2006-003-01NCVThe inspectors identified a Severity Level IV Non-CitedViolation associated with the failure to perform an

adequate safety evaluation as required by 10 CR 50.59 for

changes made to the facility as described in the UFSAR.

LIST OF DOCUMENTS REVIEWED

References for Section 1R04Procedure 2.2.19, Residual Heat Removal (RHR) SystemP&ID M241, Residual Heat Removal