Text

August 1, 2006

Mr. 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-35

Enclosure:

Inspection Report 50-293/06-03

w/Attachment:

Supplemental Information

SUMMARY OF FINDINGS

..................................................iii

REPORT DETAILS

........................................................1

REACTOR SAFETY

.......................................................11R04Equipment Alignment...........................................1

1R05 Fire Protection................................................2

1R11 Licensed Operator Requalification.................................31R12Maintenance Rule .............................................41R13Maintenance Risk Assessments and Emergent Work Control............41R15Operability Evaluations.........................................81R19Post-Maintenance Testing ......................................81R22Surveillance Testing...........................................9

1R23 Temporary Plant Modifications..................................10

1EP4Emergency Action Level and Emergency Plan Changes ..............101EP6Drill Evaluation...............................................11RADIATION SAFETY .....................................................122PS1Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems

OTHER ACTIVITIES

[OA]..................................................144OA1Performance Indicator Verification ...............................14 4OA2Identification and Resolution of Problems .........................154OA3Event Follow-up..............................................17 4OA5Other......................................................17 4OA6Meetings, Including Exit........................................18

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

..............................................A-1

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED.........................A-1

LIST OF DOCUMENTS REVIEWED

........................................A-2

LIST OF ACRONYMS

....................................................A-6

EnclosureiiiSUMMARY

OF [[]]

FINDINGSIR 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 FindingsCornerstone: 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

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

UFS [[]]

AR 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 ViolationsNone.

REPORT [[]]

DETAILSSummary 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

SAFET [[]]

YCornerstones: 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.FindingsNo 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. FindingsNo 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 drill 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.1R11Licensed 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

UFS [[]]

AR 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 tasks, 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.FindingsNo 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.

4Enclosure b.FindingsNo findings of significance were identified.1R12Maintenance 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.FindingsNo findings of significance were identified.1R13Maintenance 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.FindingsIntroduction: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

UFS [[]]

AR 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

ER 05120679 did not change a "method of evaluation" described in the

UFS [[]]

AR 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

NUR [[]]

EG0612 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

CFR 50.59 involved conditions evaluated as having very low safety significance by

the

SDP.T his 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

UFS [[]]

AR 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

UFS [[]]

AR

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

UFS [[]]

AR 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.1R15Operability 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,

UFS [[]]

AR , 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.FindingsNo findings of significance were identified.1R19Post-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 confirmed 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

MR06104965; 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.FindingsNo findings of significance were identified.1R22Surveillance 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.

10Enclosure b.FindingsNo findings of significance were identified.1R23Temporary 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. FindingsNo 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.

11Enclosure b.FindingsNo 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 notification system, and appropriately completed and transmitted the eventnotification forms in a timely manner. c.FindingsNo findings of significance were identified..2Combined Functional Drill a.Inspection Scope (1 sample) The inspectors reviewed the combined functional drill 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 June 1, 2006. b.FindingsNo findings of significance were identified.

2Enclosure2.RADIATION

SAFETY Cornerstone: Public Radiation Safety2

PS1Radioactive 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

NUR [[]]

EG-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

OD [[]]

CM 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

UFS [[]]

AR, 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 cleaning 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

UFS [[]]

AR 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.T he 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.FindingsOne 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

PN [[]]

PS 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. 4.

OTHER [[]]

ACTIVI [[]]

TIES [[[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.FindingsNo 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

2PS 1 and 4

OA1 of the

report attachment. b. FindingsNo findings of significance were identified.4OA2Identification and Resolution of Problems (71152)1.Routine Review of Corrective Action Program Issues a.Inspection ScopeAs 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.FindingsNo findings of significance were identified. 2.Corrective Action Program Semi-annual Trend Review a.Inspection ScopeAs 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 4th quarter 2005 corrective action trend reports,and the daily plant status report listings of operations equipment problems, operabilityevaluations, and temporary alterations. b. Findings and ObservationsNo 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

CR s, 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. .3Annual 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 ScopeThe 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.FindingsNo 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.

18Enclosure b.FindingsNo 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:

SUPPLE [[]]

MENTAL [[]]

INFORM [[]]

ATION A-1AttachmentSUPPLEMENTAL

INFORM [[]]

ATIONK EY

POINTS [[]]

OF [[]]

CONTAC [[]]

TLicensee 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 InspectorC. Welch, Resident InspectorLIST

OF [[]]

ITEMS [[]]

OPENED ,

CLOSED [[]]

AND [[]]

DISCUSSEDOpened05000293/2006003-02URIAnisokinetic 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 Closed05000293/2006-003-01NCVThe 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 [[]]

DOCUME NTS

REVIEW [[]]

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

P& [[]]

ID M219, Diesel Generator Air Start System

P& [[]]

ID M223, Diesel Oil Storage and Transfer System

Procedure 2.2.8, Standby AC Power System (Diesel Generators)

Procedure 8.9.1, Emergency Diesel Generator and Associated Emergency Bus SurveillanceReferences for Section 1R05Procedure 1.4.23, Fire Brigade Training DrillCondition Reports 2006-1751, 2006-01758, 2006-01759

Fire Hazards Analysis (Fire Zones 2.1, 4.1, 4.2)

5.5.2, Special Fire Fighting ProcedureReferences for Section 1R11UFSAR Section 10.3.6, Consequences of a Dropped Fuel CaskSafety Evaluation

SE 3402, Revision 1 for

ER 05120679 dated 5/10/06

ER 05120679, Use of Duratek Supplied

CNS -55 Shipping Cask

EP -

IP-100, Emergency Classification and Notification

Emergency Event notification Form 20060522002

Emergency Operating Procedures

EOP -1,

RPV Control

Emergency Operating Procedures EOP -3, Primary Containment Control

Emergency Operating Procedures EOP -5, Radioactivity Release Control

Emergency Operating Procedures

EOP -17, Emergency

RPV Depressurization

Procedure 2.1.6, Reactor Scram, Revision 58

Procedures 2.1.14, 2.2.46, 2.4.36, and 5.5.3Condition Reports 2006-01917, 200601923References for Section 1R12PNPS Maintenance Rule (a)(1) Systems StatusPilgrim System Health Reports Systems 50, 54, 66 1st Quarter 2006Top Ten Action Plan - System Backlogs, Update 5/1/2006

ENN -

DC -171, Maintenance Rule Monitoring

ENN -

DC-121, Maintenance Rule

Maintenance Rule System Structure Component (SSC) Basis Document - NE16.03,

System 54 Reactor Vessel Maintenance and Condition Reports 2005-2006

System 50 Primary Containment Maintenance and Condition Reports 2005-2006

A-3AttachmentSystem 66 C19A/B System Maintenance and Condition Reports 2005-2006Condition Reports 200601272, 200503299, 200601442

OD [[]]

MI Action Plan Unidentified Drywell LeakageReferences for Section 1R13Maintenance Request 05118433, Clean Spent Fuel Pool, Ship Irradiated Hardware6 Day Cask Handling Plan

Temporary Procedure

TP 06-011, Handling Procedure for the Duratek Transport Cask

CNS 3-55, Certificate fo Compliance 5805Duratek Procedure

TR -

OP -019, Handling Procedure for the Duratek Transport Cask

CNS 3-55,Certificate of Compliance 5805Procedure 3.M.1-14, General Maintenance Procedure for Heavy Load Handling Operations

UFSAR Section 10.3.6, Consequences of a Dropped Fuel Cask

Safety Evaluation

SE 3402, Revision 1 for

ER 05120679 dated 5/10/06

ER 05120679, Use of Duratek Supplied

CNS-55 Shipping Cask, Crusher Shear Tool/Stand andassociated Rigging for Spent Fuel Pool Cleanup Project Activities dated 4/24/06ER 05120679 5059 Screen Evaluation dated March 24, 2006

ER 05120679 5059 Screen Evaluation Updated (white paper) dated April 11, 2006

Procedure 1.3.34.15, Protected Area Postings

Procedure 1.5.22, Risk Assessment Process

Procedure 2.1.12.2, Station Blackout Diesel Generator Daily SurveillanceScheduler's Evaluation for "A" EDG Overhaul during the week of 6/11/06

Condition Reports 200601809, 200601824, 200602460

Radiation Work Permit 2006-072

Pilgrim License Application Amendments #20, #24, #29 dated 3/29/71NRC Generic Letter 78-17 dated 5/17/78, Control of Heavy Loads Near Spent FuelNRC

RIS 2005-25, Clarification of

NRC [[Guidelines for Control of Heavy LoadsNRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the ReactorCore, or Over Safety Related EquipmentNUREG 0612, Control of Heavy LoadsNRC Generic Letter dated 12, 22, 1980, Control of Heavy LoadsBECo Letter 81-242,]]

NUREG 0612 Control of heavy Loads

BEC o Letter 81-141 dated 6/25/81,

NUREG 0612 Control of Heavy Loads

BEC o Letter 83-181 dated 7/13/83,

NUREG 0612 Control of Heavy Loads (Enclosures 1 and 2)

NRC Letter dated 3/6/85, Control of Heavy Loads (Phase I) and Safety EvaluationFranklin Research Center Technical Evaluation Report dated 1/31/85, Control of Heavy LoadsBECO Response to Bulletin 96-02, Letter 96-053 dated 5/28/96

BEC o Letter #78-109 dated 6/26/78, Cask Handling Evolutions Associated with the Spent FuelPool Modification

BEC o Letter #78-123 dated 7/17/78, Response to Request for Information on Movement ofHeavy Loads Near Spent FuelLicense Amendments #33 dated 8/17/1978; #155 dated 6/22/94;

BEC o Letter dated 2/11/93, Proposed Technical Specification Change

PNPS Spent FuelStorage Capacity ExpansionHoltec Report HI-93971 dated 3/4/93, Analysis of Cask Drop in Pilgrim Spent Fuel Pool toEstablish Maximum Cask Allowable WeightDrawings A709, M20

A-4AttachmentCalculation C15.0.3445, use of Duretek Supplied

CNS 3-55 and

CNS 8-120 Shipping Casks,5.5 Ton Transfer Bell and the Crusher Shear Tool/stand in the Spent Fuel Pool CleanupProject dated 4/20/06Radiological Survey Form Map #80 dated 5/5/6, Disposal Liner Survey (under water)

Radioactive Material Shipment Truck Survey From dated 5/17/06, Shipment 06-04 - OutgoingSurvey of

SFP Project WasteReferences for Section 1R15

CR 200601849, Core Shroud Repair Tie Rod Upper Support CrackingGE Safety Communication SC06-07, Core Shroud Repair Tie Rod Upper Support Cracking

Drawing M1B51, Sheet 2 of 4, Reactor Modification Shroud Repair

Drawing M1B53, Stabilizer Support Assembly Shroud Repair

GE Safety Communication

SC06-01, Single Failure Suppression Pool Temperature Analysis

Condition Report and OE 20060254

Standing Order 06-06, Suppression Pool Temperature Post

LO [[]]

CA with loss of B17 or B18

MR 06107824, Inspection of Salt Service Water Pump P208E on 6/1/06

EOOS Risk assessment dated 6/1/06 for proposed

SSW configuration

Maintenance procedure 3.M.4-85, Station Diving Procedure for Underwater Work andInspectionsTagout 1-Cycle-16-04887 and 04904

1.3.34.5, Protected Areas on 6/1/6Technical Specification 3.5.B.4References for Section 1R19MR05117438, Seismic Monitor Event Indicator and Recorder Did Not Function When Tested3.M.3-51, Electrical Termination Procedure

Kinemetrics Inc Form #343083, Channel Calibration of Strong Motion Time History AccelerationRecorder

SMA -3/

SMP-1, 3/30/06UFSAR Section 12.2.3.5.2, Seismic Instrumentation

License Amendment 20 dated 2/11/71

Safety guide 12, Instrumentation for Earthquakes

Design Basis Document

TD [[]]

BD -11, Seismic Design

PD [[]]

CR 78-24, Seismic Monitoring Instrumentation

Emergency Procedure 5.2.1, Earthquakes, Revision 24

Alarm Response Procedure ARP-C903R-B1

Condition Reports 200403582, 20050720, 200504278 and 200504998

Specification E576,

PNPS Seismic InstrumentationReferences for Section 1R23

TA-06-1-08, Temporary Power to Neutron and Process Radiation Monitoring InstrumentationUFSAR Section 8.7 24 Volt DC Power System

Drawing E14, Schematic Diagram Vital and Radiation Protection AC System

Temporary Procedure

TP 06-015, Temporary 24 V

DC Power Feed for the 24V

DC SystemDuring Battery Testing/Replacement Tags for

TA-06-1-08

MR 05119395

A-5AttachmentReferences for Section

2PS 1 and 4

OA1Procedure 7.3.25, Rev 33, Particulate and Iodine Monitoring at the Main Stack and ReactorBuilding VentProcedure 7.3.31, Rev. 17, Tritium Sampling

Procedure 7.3.36, Rev. 49, Offgas Sample Analysis

Procedure 7.3.37, Rev. 31, Determination of Conversion Factors for Gaseous PRMs

Procedure 7.3.48, Rev. 8, Airborne Effluent Monitoring of the turbine Deck and Reactor FeedPump BayProcedure 7.4.12, Rev. 22, Calibration of the

SJ [[]]

AE Offgas Process Rad Monitors

Procedure 7.4.42, Rev. 23, Calibration of the

NUMAC Gaseous

PRMs

Procedure 7.4.48, Rev. 4, Calibration of Turbine Building Gaseous Effluent Monitors (GEMS)

Procedure 7.4.49, Rev. 4, Operation of Turbine Building Gaseous Effluent Monitors (GEMS)Procedure 7.4.63, Rev. 2, Process Rad Monitor Setpoints

Procedure 7.4.64, Rev. 2, Process Radiation Monitor Alarm Response

Procedure 7.4.42, Rev. 19, Process Radiation Monitor Calibrations

Procedure 7.8.13, Rev 2, Chemistry Actions During Plant Transients

Procedure 7.8.1, Rev. 41, Chemistry Sample and Analysis Program

Procedure 7.9.1, Retired, Gaseous Waste Discharge Procedure

Procedure 7.9.12, Rev. 3, Liquid Radwaste Verification and Discharge

Procedure 7.9.15, Rev. 0, Dose Assessment (2PS1 and 4OA1)

Procedure 7.10.3, Rev 17,

PRM Cal Check (and test data generated 6/14/06)Procedure

EN-WM-100, Rev. 0, Work Request Generation, Screening, and ImplementationCR 2006-00134, 1/12/06, Augmented Offgas System out of service longer than necessary.CR 2006-01059, 1/20/06, Procedural conflicts

CR 2006-00400, 1/31/06, Water management and sump discharge

CR 2006-02282, 6/15/06, Isokinetic sampling of Turbine Building Vent and Main Stack

CR 2006-02266, 6/14/06,

ODCM does not include Turbine Building effluent monitoring systemSelf-Assessment, Chemistry

QC and Instrument Performance, Oct 26-27, 2004

Snapshot assessment for effluent dose, 6/22/05

Manager's Focused Assessment of Chemistry Instruments, 9/25/05

Pilgrim Nuclear Power Station Radiological Effluent and Waste Disposal Report, January 1through December 31, 2005 (2PS1 and

4OA 1)Pilgrim Nuclear Power Station Radiological Effluent and Waste Disposal Report, January 1through December 31, 2004 (2

PS 1 and

4OA 1)Calibration/Testing Procedure Records 7.4.42, Rev. 19, Process Radiation Monitor Calibrations,1/22/04, 3/10/05, 3/17/05, 6/16/05Calibration/Testing Procedure Records 3.M.2-6.4, Rev. 17,

NUMAC Process Radiation MonitorCalibrationCalibration/Testing Procedure Records 8.E.8, Rev. 37, Offgas Instrument Calibration(electrical), 6/7/06, 1/23/06Calibration/Testing Procedure Records 8.F.8, Rev 12, Offgas system Instruments Calibration,3/16/04Calibration/Testing Procedure Records 8.M.3-9, Rev. 24, Liquid Radwaste Effluent DischargeMonitor Functional Test, 4/26/06Calibration/Testing Procedure Records 8.M.3-17.1, Rev. 6, Radioactive Liquid EffluentAlternate Flow Rate (Liquid Level) Instrument Functional and Calibration, 1/30/06

A-6AttachmentCalibration/Testing Procedure Records

8.M. 2-4.1, Rev. 31, Air Ejector Offgas Log RadiationMonitor CalibrationReferences for Section 4

OA21.3.34.4, "Compensatory Measures," Rev. 14References for Section 40A51.3.12, "Notification and Recall of Personnel", Rev 391.5.22, "Risk Assessment Process", Rev 8

2.1.15, "Daily Surveillance Log", Rev 178

2.4.16, "Distribution Alignment Electrical System Malfunctions", Rev 31

2.4.144, "Degraded Voltage", Rev 32

5.3.31, "Station Blackout", Rev 10

8.C. 34, "Operations Technical Specifications Requirements for InoperableSystems/Components", Rev 37

EN -WM-101, "On-line Work Management Process", Rev

0LIST [[]]

OF [[]]

ACRONY [[]]

MSADAMSAgencywide Documents Access and Management SystemCFRCode of Federal Regulations

CRC ondition Report

ED [[]]