IR 05000354/2007002

IR 05000354-07-002, on 01/01/2007 - 03/31/2007; Hope Creek Generating Station; Resident Inspector Integrated Report

ML071290407
Person / Time
Site:	Hope Creek
Issue date:	05/09/2007
From:	Burritt A L Reactor Projects Branch 3
To:	Levis W Public Service Enterprise Group
BURRITT, AL
References
IR-07-002
Download: ML071290407 (43)

Text

May 9, 2007

Mr. William LevisPresident and Chief Nuclear Officer PSEG LLC - N09 P. O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT: HOPE CREEK GENERATING STATION - NRC INTEGRATED INSPECTIONREPORT 05000354/2007002

Dear Mr. Levis:

On March 31, 2007, the US Nuclear Regulatory Commission (NRC) completed an inspection atyour Hope Creek Generating Station. The enclosed integrated inspection report documents the inspection results, which were discussed on April 5, 2007, with Mr. George Barnes and other 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, observed activities, and interviewed personnel.Based on the results of this inspection, no findings of significance were identified. However,licensee-identified violations which were determined to be of very low safety significance are listed in this report. The NRC is treating these violations as non-cited violations (NCVs)

consistent with Section VI.A.1 of the NRC Enforcement Policy because of the very low safety significance of the violations and because they are entered into your corrective action program.

If you contest these non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Hope Creek Generating Station.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the Mr. W. Levis2 2NRC Public Document Room or from the Publicly Available Records (PARS) component ofNRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Arthur L. Burritt, ChiefProjects Branch 3 Division of Reactor ProjectsDocket No:50-354License No:NPF-57

Enclosure:

Inspection Report 05000354/2007002

w/Attachment:

Supplemental Informationcc w/encl:G. Barnes, Site Vice President D. Winchester, Vice President - Nuclear Assessments B. Clark, Director - Finance J. Perry, Hope Creek Plant Manager J. J. Keenan, General Solicitor, PSEG M. Wetterhahn, Esquire, Winston and Strawn, LLP Consumer Advocate, Office of Consumer Advocate, Commonwealth of Pennsylvania L. A. Peterson, Chief of Police and Emergency Management Coordinator P. Baldauf, Assistant Director of Radiation Protection Programs, State of New Jersey K. Tosch, Chief, Bureau of Nuclear Engineering, NJ Dept. of Environmental Protection

H. Otto, Ph.D., Administrator, Interagency Programs, DNREC Division of Water Resources, State of Delaware N. Cohen, Coordinator - Unplug Salem Campaign E. Zobian, Coordinator - Jersey Shore Anti Nuclear Alliance M

SUMMARY OF FINDINGS

IR 05000354/2007002; 01/01/2007 - 03/31/2007; Hope Creek Generating Station; ResidentInspector Integrated Report.The report covered a 13-week period of inspection by resident inspectors, regional healthphysicist inspectors, and regional reactor inspectors. No findings of significance were identified.

The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.A.

NRC-Identified and Self-Revealing Findings

No findings of significance were identified.

B.Licensee Identified Violations

Violations of very low safety significance, that were identified by Public Service EnterpriseGroup (PSEG) have been reviewed by the inspectors. Corrective actions taken or planned by PSEG have been entered into PSEG's corrective action program. These violations and corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant StatusThe Hope Creek Generating Station began the first quarter operating at 100% power. The plantwas shutdown on January 26, 2007, to cold shutdown conditions to execute a scheduled maintenance outage.During power ascension on January 29, 2007, Hope Creek automatically scrammed on lowreactor water level caused by a failed reactor feed pump minimum flow valve. The plant was returned to 100% power on February 2, 2007, and remained at or near full power for the remainder of the inspection period.1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01)

a. Inspection Scope

(1 sample)The inspectors reviewed seasonal adverse weather preparation activities related to rivergrass intrusion conditions that impact the station service water system. Inspectors assessed implementation of PSEG's grassing readiness plan through plant walkdowns, corrective action program review, and discussions with cognizant managers and engineers. Documents reviewed by inspectors are listed in the attachment.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04).1Partial Walkdown (5 samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems to verify theoperability of redundant or diverse trains and components when safety equipment was inoperable. The inspectors completed walkdowns to identify any discrepancies that could impact the function of the system, and therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, walked down control systems components, and verified that selected breakers, valves, and support equipment were in the correct position to support system operation. The inspectors also verified that PSEG had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program. Documents reviewed are listed in the attachment.

2Enclosure 'A' and 'C' 1E switchgear during planned maintenance on the 'B' 1E switchgear Redundant station service water (SSW) trains and support equipment duringmaintenance on the 'D' SSW pump and traveling water screen The 'A' & 'C' SSW trains, emergency diesel generators (EDGs), and 4KVswitchgear rooms during the emergent unavailability of the 'B' & 'D' SSW trains Redundant EDG, emergency core cooling systems (ECCS), SSW, filtration,recirculation, and ventilation system (FRVS), station auxiliary cooling system (SACS), and control room (CR) chilled water equipment during extended planned maintenance on the 'C' EDG and unplanned unavailability of the 'B' FRVS vent fan and 'A' CR chiller 'B' control room chilled water system after return to service following extendedplanned maintenance

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05).1Fire Protection - Quarterly Tours

a. Inspection Scope

(10 samples)The inspectors conducted tours of ten areas to assess the material condition andoperational status of fire protection features. The inspectors verified that combustible material and ignition sources were controlled in accordance with PSEG's administrative procedures; fire detection and suppression equipment was available for use; that passive fire barriers were maintained in good material condition; and that compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with PSEG's fire plan. The areas toured are listed below with their associated pre-fire plan designator. Other documents reviewed are listed in the attachment.FRH-II-571, diesel area heating, ventilation and air conditioning (HVAC)equipment roomFRH-II-563, control area HVAC equipment roomsFRH-II-552, control room areaFRH-III-133, accessible turbine building rooms containing offsite power sourcebus ducts to safety-related 4KV bussesFRH-II-412, 'D' residual heat removal (RHR) pump and reactor core isolationcooling (RCIC) pump roomsFRH-II-531, Common EDG Corridor, 102' ElevationFRH-II-471, Refuel Floor, 201' ElevationFRH-II-424, Motor Control Center (MCC) Area, Room 4218, 77' ElevationFRH-II-431, MCC Area, Room 4303, 102' ElevationFRH-II-151, 'A' Recirc MG Set Room, 137' Elevation

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06).1Internal Flooding

a. Inspection Scope

(1 sample)The inspectors reviewed selected risk-important plant design features and PSEGprocedures intended to protect the plant and its safety-related equipment from internal flooding events. The inspectors focused on mitigation strategies and equipment in the 'B' RHR pump room. The inspectors reviewed flood analysis and design documents, including the updated final safety analysis report, engineering calculations, and abnormal operating procedures. The inspectors observed the condition of wall penetrations, watertight doors, flood alarm switches, and drains to assess their readiness to contain flow from an internal flood in accordance with the design basis. In addition, the inspectors walked down the 'B' RHR room and adjacent rooms in the reactor building to assess potential flooding vulnerabilities.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11).1Requalification Activities Review By Resident Staff

a. Inspection Scope

(1 sample)The resident inspectors observed one annual licensed operator requalification simulatorexamination scenario on January 20, 2007, to assess operator performance and training effectiveness. The scenario involved a main turbine vibration problem, a failed reactor mode switch, and a steam leak in the turbine building. The inspectors assessed simulator fidelity and observed the simulator instructor's critique of operator performance.

The inspectors also observed control room activities with emphasis on simulator identified areas for improvement. Finally, the inspectors reviewed applicable documents associated with licensed operator requalification as listed in the attachment.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

(71111.12)

a. Inspection Scope

(3 samples)4Enclosure The inspectors reviewed the three samples listed below for items such as:

(1) appropriate work practices;

(2) identifying and addressing common cause failures; (3)scoping in accordance with 10 CFR 50.65(b) of the maintenance rule (MR);

(4) characterizing reliability issues for performance;

(5) trending key parameters for condition monitoring;

(6) charging unavailability for performance;

(7) classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); and

(8) appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified as (a)(1). Documents reviewed are listed in the attachment. Items reviewed included the following:'A' RHR pump minimum flow valve failed to close;GS-HV-5029 reactor building to suppression chamber vacuum breaker isolationvalve slow closure; and'C' reactor auxiliaries cooling system (RACS) pump motor failure.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

(5 samples)The inspectors reviewed on-line risk management evaluations through direct observationand document reviews for the following configurations:Planned maintenance on the 'A' EDG on January 3, 2007;'B' 1E switchgear relay outage testing reclassified as online and performed onFebruary 8, 2007;Concurrent planned maintenance on the 'D' SSW pump, 'A' circulating waterpump, 'B' primary containment instrument gas compressor, and the 10K107 service air compressor on February 20-22, 2007;Unplanned unavailability of the 'B' FRVS vent fan and 'A' control room chillerduring planned extended maintenance on the 'C' EDG on March 7, 2007; andEmergent unavailability of the 'B' electro-hydraulic control (EHC) pump on March27 and 28, 2007.The inspectors reviewed the applicable risk evaluations, work schedules and controlroom logs for these configurations to verify that concurrent planned and emergent maintenance and test activities did not adversely affect the plant risk already incurred with these configurations. PSEG's risk management actions were reviewed during shift turnover meetings, control room tours, and plant walkdowns. The inspectors also used PSEG's on-line risk monitor (Equipment Out-Of-Service workstation) to gain insights into the risk associated with these plant configurations. Finally, the inspectors reviewed notifications documenting problems associated with risk assessments and emergent work evaluations. Documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

(5 samples)The inspectors reviewed five operability determinations for degraded or non-conformingconditions associated with:SACS pipe support failure on December 20, 2006;Operation of the 6B feedwater heater with water level low-out-of-specification on December 31, 2006;'A' control room chiller temperature control valve inoperability on February 10, 2007;'B' SSW lube water supply system through-wall leakage on February 22 - 28,2007; and'B' FRVS vent fan unplanned inoperability on March 5, 2007.The inspectors reviewed the technical adequacy of the operability determinations toensure the conclusions were justified. The inspectors also walked down accessible equipment to corroborate the adequacy of PSEG's operability determinations.

Additionally, the inspectors reviewed other PSEG identified safety-related equipment deficiencies during this report period and assessed the adequacy of their operability screenings. Notifications and documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

a. Inspection Scope

(1 sample)The inspectors reviewed a design change associated with a valve (DA-HV-2097) in theservice water structure deicing line. The modification changed the controls of the motor operator on the valve such that the valve will be not open beyond 12% of full-open. The modification was installed to limit the amount of circulating water diverted from the cooling tower basin to the service water intake structure to minimize the chance of silt disturbance near the service water pump suctions.The design bases, licensing bases, modification instructions and post modification testingof the affected components were reviewed to verify the performance capability of this equipment was not adversely affected. The inspectors reviewed the applicable technical specifications for this equipment to ensure that operability requirements and allowable outage time limits were met. The inspectors also reviewed notifications documenting 6Enclosure deficiencies identified related to permanent plant modifications. The documentsreviewed as part of these inspections are listed in the attachment.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

(6 samples)The inspectors reviewed the post-maintenance tests listed below to verify thatprocedures and test activities ensured system operability and functional capability. The inspectors reviewed test procedures to verify the procedure adequately tested the safety functions that may have been affected by the maintenance activity and the acceptance criteria in the procedure were consistent with the Updated Final Safety Analysis Report (UFSAR) and other design or license basis documentation. The inspectors also witnessed the test or reviewed the test data to verify test results adequately demonstrated restoration of the affected safety functions. Documents reviewed are listed in the attachment.WO 60056618, 'A' emergency service water makeup valve design change*WO 60065779, 'B' H2/O2 analyzer isolation valves bailey card replacement

• WO 30139614, replacement of the 'A' reactor recirculation pump seal package

• WO 60066866, repair of steam leak on the 3A feedwater heater extraction steam piping*WO 60067170 and 60066955, repair of 'A' and 'C' drywell to suppressionchamber vacuum breaker indications*WO 60055819, 'C' EDG keepwarm pump replacementUltrasonic measurement data associated with the 3A feedwater heater extraction steampiping was reviewed by a NRC regional specialist. The repair methods and post-maintenance testing methodology was also reviewed by the regional specialist and determined to be adequate.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

.1Scheduled Maintenance Outage on January 26, 2007

a. Inspection Scope

(1 sample)The plant was shutdown on January 26, 2007, to implement a planned maintenanceoutage. The primary purpose of the outage was to repair a steam leak on an extraction 7Enclosure steam line providing steam to the 3A feedwater heater and to replace the shaft sealpackage on the 'A' reactor recirculation pump. The inspectors reviewed these maintenance activities and they are documented in section 1R19, Post-Maintenance Testing.The inspectors reviewed PSEG's outage schedule and activities to verify that risk wasconsidered appropriately and that license and technical specification requirements were adhered to. The inspectors observed portions of the reactor shutdown and subsequent start up from the control room to verify PSEG adhered to station procedures and to evaluate operator performance. The inspectors toured areas of the plant that were normally inaccessible during power operations to verify that safety related and risk significant SSCs were maintained in an operable condition. The inspectors performed a walkdown of the drywell following completion of all maintenance activities to verify there was no evidence of system leakage and that debris had not been left behind that could affect performance of plant equipment. Documents reviewed are listed in the attachment.Hope Creek completed the scheduled maintenance outage on January 29, 2007, at10:51 pm when the main generator was synchronized to the 500 kV grid. At 11:10 pm the reactor protection system automatically inserted all control rods into the reactor core due to a reactor pressure vessel (RPV) water level control problem. The transient is described in more detail in section

==4OA3 , Event Followup.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

==

a. Inspection Scope

(5 samples)The inspectors witnessed five surveillance tests and reviewed test data of selectedsurveillance tests listed below to verify that the test met the requirements of the Technical Specifications, UFSAR, and station procedures. The inspectors also determined whether the testing effectively demonstrated that the structures, systems, and components were operationally ready and capable of performing their intended safety functions. Documents reviewed are listed in the attachment.*WO 50099699, high pressure coolant injection (HPCI) system in-service test onJanuary 10, 2007*WO 50087561, 'B' emergency diesel generator 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> endurance test onJanuary 17, 2007*WO 50099736, quarterly 'B' & 'D' core spray pump in-service test on January 18, 2007*WO 50098995, control rod scram time surveillance on January 26, 2007

• WO 50101674, Class 1E, Channel D, 125 Volt Quarterly Battery Surveillance onMarch 21, 2007

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications (71111.23)

a. Inspection Scope

(1 sample)The inspectors reviewed a temporary plant modification (T-Mod 07-007) associated withthe 'A' control room chiller. The modification bypassed a thrust bearing high oil temperature switch that had failed and resulted in an unplanned trip of the chiller. The inspectors verified the modification was consistent with the design and licensing bases of the chilled water system and that the performance capability of the system was not degraded by the modification. The inspectors reviewed documents to verify PSEG followed their processes for implementing temporary modifications on plant SSCs. In addition, the inspectors verified the modified equipment alignment through control room instrumentation and plant walkdowns of accessible portions of the affected equipment.

The inspectors also reviewed notifications documenting problems associated with equipment affected by temporary modifications. Documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.Cornerstone: Emergency Preparedness (EP)1EP2Alert and Notification System (ANS) (71114.02)

a. Inspection Scope

(1 sample)An onsite review was conducted to assess the maintenance and testing of PSEG's ANS. During this inspection, the inspectors interviewed site EP staff responsible for implementation of the ANS testing and maintenance. Notifications pertaining to the ANS were reviewed for causes, trends, and corrective actions. The inspectors further discussed with PSEG the new ANS system design and its benefits over the previous system. The inspectors reviewed PSEG's original ANS design report to ensure compliance with those commitments for system maintenance and testing. The inspectors toured the Emergency Operations Facility (EOF). On March 28, 2007, the inspectors observed a silent test of the ANS. Applicable emergency planning standards of 10 CFR 50.47 and the related requirements of 10 CFR 50, Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.

9Enclosure

1EP3 Emergency Response Organization (ERO) Staffing and Augmentation System

a. Inspection Scope

(1 sample)A review of Salem/Hope Creek's ERO augmentation staffing requirements and theprocess for notifying the ERO was conducted. This was performed to ensure the readiness of key staff for responding to an event and to ensure timely facility activation.

The inspectors reviewed procedures, notifications, and call-in drills associated with the ERO notification system and drills. The inspectors interviewed personnel responsible for testing the ERO augmentation process. The inspectors compared qualification requirements to the training records for a sample of ERO members. The inspectors also verified that the EP department staff were receiving required training as specified in the emergency plan. Applicable emergency planning standards of 10 CFR 50.47 and the related requirements of 10 CFR 50, Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.1EP4Emergency Action Level (EAL) and Emergency Plan Changes (71114.04)

a. Inspection Scope

(1 sample)Prior to this inspection, the NRC had received and acknowledged changes made to theSalem/Hope Creek Emergency Plan and implementing procedures. PSEG developed these changes in accordance with 10 CFR 50.54(q), and determined that the changes did not result in a decrease in effectiveness to the Plan. PSEG also determined that the plan continued to meet the requirements of 10 CFR 50.47(b) and 10 CFR 50 Appendix E.

During this inspection, the inspectors conducted a sampling review of Salem/Hope Creek's 10 CFR 50.54(q) screenings for the changes made to the Plan that could potentially result in a decrease in effectiveness. This review did not constitute NRC approval of the changes and, as such, the changes remain subject to future NRC inspection. Also, the NRC reviewed PSEG's EAL scheme for logic and consistency. The requirements in 10 CFR 50.54(q) were used as reference criteria.

b. Findings

No findings of significance were identified.1EP5Correction of Emergency Preparedness Weaknesses (71114.05)

a. Inspection Scope

(1 sample)The inspectors reviewed EP self-assessments and audit reports to assess PSEG's abilityto evaluate their performance and programs. The inspectors reviewed notifications initiated from December, 2005 to March, 2007 at Salem/Hope Creek from drills, self-10Enclosure assessments, and audits. Applicable emergency planning standards of 10 CFR 50.47and the related requirements of 10 CFR 50, Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.1EP6Drill Evaluation (71114.06)

a. Inspection Scope

(1 sample)Resident inspectors evaluated the conduct of a simulator examination scenario onJanuary 20, 2007, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the simulated control room to verify that event classification and notifications were done in accordance with the Hope Creek Event Classification Guide. The inspectors also observed PSEG's critique of the examination to compare any inspector-observed weakness with those identified by PSEG personnel to verify whether PSEG was properly identifying weaknesses.

b. Findings

No findings of significance were identified.2.RADIATION SAFETYCornerstone: Occupational Radiation Safety2OS1Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

(7 samples)The inspectors reviewed all PSEG performance indicators for the Occupational RadiationSafety Cornerstone for followup.The inspectors identified exposure significant work areas within radiation areas, highradiation areas (<1 R/hr), or airborne radioactivity areas in the plant and reviewed associated PSEG controls and surveys of these areas to determine if controls (e.g.

surveys, postings, barricades) were acceptable.The inspectors walked down these areas or their perimeters to determine: whetherprescribed radiation work permits, procedure, and engineering controls were in place, whether PSEG surveys and postings were complete and accurate, and whether air samplers were properly located.

11Enclosure The inspectors examined PSEG's physical and programmatic controls for highlyactivated or contaminated materials (non-fuel) stored within spent fuel and other storage

pools.The inspectors discussed with the Radiation Protection Manager high dose rate - highradiation areas, and very high radiation areas (VHRA) controls and procedures. The inspectors verified that any changes to PSEG procedures do not substantially reduce the effectiveness and level of worker protection.The inspectors discussed with first-line health physics supervisors the controls in placefor special areas that have the potential to become VHRA during certain plant operations.The inspectors reviewed and assessed the adequacy of PSEG's internal doseassessment for any actual internal exposure greater than 50 mrem committed effective dose equivalent (CEDE).

b. Findings

No findings of significance were identified.2OS2ALARA Planning and Controls (71121.02)

a. Inspection Scope

(2 samples)The inspectors reviewed the assumptions and basis for the current annual collectiveexposure estimate. The inspectors reviewed applicable procedures to determine the methodology for estimating work activity-specific exposures and the intended dose outcome.The inspectors reviewed the exposure results and monitoring controls of declaredpregnant workers. A total of six personnel were declared pregnant workers during 2006, with the maximum dose to an individual during the declaration period being 3 millirem.

b. Findings

No findings of significance were identified.2OS3Radiation Monitoring Instrumentation and Protective Equipment (71121.03)

a. Inspection Scope

(1 sample)The inspectors identified the types of portable radiation detection instrumentation usedfor job coverage of high radiation area work, other temporary area radiation monitors currently used in the plant, and continuous air monitors associated with jobs with the potential for workers to receive 50 mrem CEDE.

b. Findings

12Enclosure No findings of significance were identified.4.OTHER ACTIVITIES4OA1Performance Indicator Verification (71151)

a. Inspection Scope

(6 samples)Cornerstone: Initiating EventsThe inspectors reviewed PSEG's program to gather, evaluate and report information onthe following performance indicators (PIs). The inspectors used the guidance contained in (Nuclear Energy Institute) NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 4, to assess the accuracy of PSEG's collection and reporting of PI data. The documents reviewed by the inspectors are listed in the attachment.Unplanned SCRAMS per 7,000 Critical HoursUnplanned SCRAMS with Loss of Normal Heat RemovalUnplanned Power Changes per 7,000 Critical HoursThe inspectors verified the accuracy and completeness of reported manual andautomatic unplanned scrams during the period of January 1, 2006 through December 31, 2006, for the "Unplanned Scrams per 7,000 Critical Hours" PI.The inspectors reviewed and verified PSEG's basis for including or excluding anyunplanned reactor scrams for the "Unplanned Scrams with Loss of Normal Heat Removal" PI during the period of January 1, 2006 through December 31, 2006.The inspectors verified the accuracy and completeness of reported transients thatresulted in unplanned changes in reactor power of greater than 20 percent power for the "Unplanned Power Changes per 7,000 Critical Hours" PI during the period of January 1, 2006 through December 31, 2006.Cornerstone: Emergency Preparedness (3 samples)Drill and Exercise PerformanceERO Drill ParticipationAlert and Notification System ReliabilityThe inspectors reviewed supporting documentation from EP drills and ANS tests duringthe period of January 1, 2006 through December 31, 2006 to verify the accuracy of the reported data.

b. Findings

No findings of significance were identified.

13Enclosure

4OA2 Identification and Resolution of Problems

.1Review of Items Entered into the Corrective Action ProgramAs required by Inspection Procedure 71152, Identification and Resolution of Problems,and to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into PSEG's corrective action program. This was accomplished by reviewing the description of each new notification and attending daily management review committee meetings.

Documents reviewed are listed in the attachment..2Annual Sample: Bailey Logic Module Failures

a. Inspection Scope

(1 sample)The inspectors reviewed PSEG's actions to address an adverse trend in Bailey LogicModule failures. A number of issues have been identified in the PSEG corrective action program (CAP) describing a rising number of Bailey logic card failures. The issues were selected for review based on their potential to increase the likelihood of an initiating event or cause the inoperability of a safety system. The inspectors reviewed PSEG procedures, vendor documents, design change packages, notifications, orders, corrective actions, and apparent cause evaluations to understand the equipment functions and operational history, as well as the identification, evaluation, and corrective actions associated with the degraded conditions. System engineers, reactor operators and other PSEG staff were interviewed to gain additional insights on the failures.The following examples illustrate a sampling of issues associated with Bailey LogicModule failures:On August 25, 2006, a Bailey Logic Module failure resulted in automatic closure of aturbine auxiliaries cooling system (TACS) return isolation valve and an unplanned power reduction to 78% power. Operators manually isolated the 'B' SACS loop then restored TACS cooling to stabilize the plant. Post-event review identified that the Field Programmable Logic Array chip in the Bailey Logic Module failed causing the automatic isolation of the valve. Corrective actions included replacement of the failed logic card, failure analysis of the faulty card, and continuation of the Bailey card replacement project.On November 27, 2006, indication was lost for the 'D' emergency diesel generator outputbreaker. Operators ordered the emergent replacement of the card even though they suspected the problem only impacted the indication portion of the Bailey Logic Module.

The post-replacement testing revealed that the failure would have prevented automatic and main control room operation of the diesel output breaker. Corrective actions included replacement of the failed logic card, a more detailed failure analysis of the faulty card, and re-evaluation of the Bailey card replacement project.

b. Findings

& Observations 14Enclosure No findings of significance were identified.The inspectors found that PSEG appropriately identified degraded conditions associatedwith Bailey Logic Module failures and entered them into the corrective action program.

Evaluations of degraded conditions were thorough, and included considerations for extent of condition. The inspectors reviewed the above examples and determined that performance deficiencies did not exist. Corrective actions developed by PSEG were appropriate to adequately address identified deficiencies.4OA3Event Followup (71153).1Hope Creek Automatic Scram on January 29, 2007

a. Inspection Scope

(1 sample)Hope Creek completed a scheduled maintenance outage on January 29, 2007, at 10:51pm when the main generator was synchronized to the 500 kV grid. At 11:02 pm, control room operators observed reactor water level lower than expected and took action to restore level. Efforts to restore reactor water level were unsuccessful. At 11:10 pm, RPV water level was below 12.5 inches and the reactor protection system automatically inserted all control rods into the reactor core (a reactor scram).The inspectors responded to the control room following the reactor scram to observepost-scram operations. The inspectors collected data from the plant computer to evaluate plant conditions prior to, during, and following the transient. The inspectors observed and participated in interviews with control room operators to gain an understanding of how operators responded to the transient. The inspectors observed engineering technical analysis and evaluation meetings and interviewed engineers to gain an understanding of the transient and to assess PSEG's evaluation process. The inspectors observed the Plant Oversight Review Committee meeting prior to plant startup to evaluate whether PSEG appropriately resolved the issues that led to the transient.A root cause evaluation identified a failed reactor feed pump minimum flow valve as thecause of the level control problem and subsequent reactor scram. Corrective actions included repair of a reactor feed pump flow instrument tubing line and clarification to the low power operating portion of the feedwater system operating procedure. Documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified..2(Closed) LER 05000354/2006-005-00, Drywell Hoist Breaker Not Open Prior to Mode 2EntryOn December 18, 2006, PSEG identified that breaker 52-263042 for the drywell safetyrelief valve (SRV) hoist was in the closed position. Technical Specification 3.8.4.1 15Enclosure requires that breaker 52-263042 be administratively maintained open in OperationalCondition 1, 2, and 3. The breaker was not open and administratively controlled prior to entry into Operating Condition 2 on May 2, 2006. The inspectors reviewed the licensee event report (LER) and evaluations associated with the performance deficiency. The enforcement aspects of this finding are discussed in Section

4OA7 . This LER is closed.4OA5Other Activities.1Independent Spent Fuel Storage Installation (ISFSI)

a. Inspection Scope

(1 sample)The inspection was a follow-up to Inspection Report 05000354/2006010, completed onNovember 9, 2006. This inspection consisted of evaluating post dry cask storage activities associated with the recent completion of Hope Creek's initial ISFSI fuel loading campaign. Inspection activities consisted of interviews with cognizant personnel and reviews of PSEG documentation. Areas inspected included review of final dose totals for the initial ISFSI campaign, completed work packages, effectiveness of corrective actions implemented after loading of the first canister, PSEG identified lessons-learned during the initial campaign, ISFSI-related notifications, and verification of personnel training and qualifications.The inspectors reviewed the completed work package for the loading of the first canister. The work package included the procedures for loading and sealing the multi purpose canister (MPC), weld data sheets, liquid penetrant examination reports, and daily polar crane check lists. The inspectors verified that procedure steps were completed and necessary signatures and approvals obtained as required.The inspectors interviewed cognizant personnel regarding the meaning and purpose ofvarious signature completion steps in procedure NC.MD-PM.DCS-0003, "Sealing, Drying, and Backfilling of a Loaded MPC." PSEG personnel confirmed that signatures for various steps signified that work was successfully completed and that associated data sheets had been reviewed by qualified individuals. The inspectors discussed the training and qualification requirements for the CaskLoading Supervisor position with cognizant personnel. The inspectors determined that designated individuals were qualified as Cask Loading Supervisors in accordance with PSEG's program to meet the requirements of ANSI/ANS-3.1-1981, section 4.3.2. In addition these individuals were required to attend various training classes that includedsuch topics as contract management, supplemental personnel oversight, and QA orientation. The inspectors observed that PSEG formally documented that individuals were properly qualified per ANSI/ANS-3.1, verified that the training database contained the documentation in the records for three individuals designated as Cask Loading Supervisors, and verified the required training for these individuals was maintained current.

16Enclosure The inspectors reviewed PSEG actions in response to exceeding the first fuel campaigndose estimate by approximately 1.5 rem. PSEG conducted a post-job critique after the first loading.The inspectors reviewed PSEG's ISFSI-related corrective action notifications, lessons-learned documentation and action plans. Areas identified for evaluation included polarcrane reliability improvements, dose reduction efforts, and transporter maintenance.

b. Findings

No findings of significance were identified..2(Closed) URI 05000354/2006015-03, Inspection of PRA Quality Issues, and NRC Reviewof Human Error Probability (HEP) Assigned Value for Battery Charger Cross-tie OperatorActionThe Senior Reactor Analyst (SRA) reviewed the unresolved item (URI) crediting operatoraction, NR-XTIE-CHARGE, in PSEG's probabilistic risk assessment (PRA) to determine if the model was reasonably representative of the as-built, as-operated nuclear power unit which it represents. This action is credited during postulated loss-of-offsite power (LOOP) events with an assumed concurrent failure of the 'B' and 'D' EDGs. Specifically, the Hope Creek PRA model credits manual operator actions to cross-tie power to the 'B' or 'D' battery chargers to provide power to the SRVs for reactor pressure control. For the NR-XTIE-CHARGE action to be successful, it must be completed prior to the depletion of the batteries. During the Component Design Bases Inspection, completed on December 7, 2006, the inspection team questioned the appropriateness of assigning a HEP of 0.6 when the performance shaping factors, such as training, availability of equipment and diagnosis were not favorable. An inappropriate assignment of an HEP could have an adverse impact on the ability of PSEG to assess and manage risk during normal plant operations.The SRA conducted a sensitivity analysis for maintenance rule risk assessments whichwould be impacted by the NR-XTIE-CHARGE operator action. The most limiting maintenance configuration would be a HPCI system unavailability with 'B' SACS in standby, 'D' SSW pump in standby, 'B' control rod drive pump in standby, and air compressor 10K107 in standby. A bounding unavailability of 14 days was assigned, which is the technical specification allowed outage time for HPCI. This was consideredconservative because the 2002 - 2004 unavailability data from the Mitigating System Performance Index (MSPI) bases documents shows that the largest HPCI unavailability for this period occurred in July 2004 in which 154.23 hrs. (6.43 days). There were no significant outages of other monitored components during this period. Other MSPI components included in the sensitivity analysis were also reviewed and found to have a negligible impact. The HEP evaluated for the analysis ranged from PSEG's assigned value of 0.6 (60% chance of failure) to 1.0 (100% chance of failure). Utilizing the guidance provided in Inspection Manual Chapter (IMC) 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," the SRA concluded that under the most bounding assumptions, the changes in the maintenance 17Enclosure rule risk assessment would not be significant and would be within the acceptable rangeof PRA uncertainty.The generic issues associated with PRA quality are being addressed by the staff asoutlined and updated in SECY-07-0042, "Status of the Plan for the Implementation of the Commission's Phased Approach to Probabilistic Risk Assessment Quality," issued on March 7, 2007. The SRA concluded that the sensitivity of NR-XTIE-CHARGE did not impact the ability of PSEG to assess and manage risk during normal plant operations.

As such, for this issue it was determined that the model was reasonably representative of the as-built, as-operated nuclear power unit which it represents and this URI is closed.4OA6Meetings, Including ExitOn April 5, 2007, the inspectors presented their findings to members of PSEG management led by Messrs. Barnes and Perry. None of the information reviewed by the inspectors was considered proprietary. 4OA7Licensee-Identified ViolationsThe following violations of very low safety significance (Green) were identified by PSEGand are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as Non-Cited Violations.Technical Specification 6.12.1 requires that access to, and activities in, each highradiation area be controlled by means of a radiation work permit. Entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. On December 12, 2006, a mechanical maintenance technician ascended into the Unit 1 turbine building crane cab, a posted high radiation area, without having been briefed or signed in on the appropriate high radiation area radiation work permit. The event is documented in PSEG's CAP as notification 20306791. The finding is only of very low safety significance because it did not involve a very high radiation area or personnel over-exposure.Technical Specification 3.8.4.1, "Primary Containment Penetration ConductorOvercurrent Protective Devices," requires that breaker 52-263042 for the drywell SRV hoist be administratively maintained open in Operational Conditions 1, 2, and 3. Contrary to this requirement, on December 18, 2006, PSEG identified that this breaker was closed. PSEG entered this issue into their corrective action program as notification 20307894. It was subsequently determined that the breaker had been closed since Hope Creek entered Operating Condition 2 on May 2, 2006. The issue was determined to be of very low safety significance, based on IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, because the finding does not represent an actual open pathway in the physical integrity of reactor containment.

18Enclosure ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

G. Barnes, Station Vice President

M. Massaro, Plant Manager

J. Perry, Plant Manager

D. Burgin, Emergency Preparedness Manager

D. Kelly, Radiation Protection Technical Support Manager

B. Sebastian, Radiation Protection Manager

B. Booth, Operations Director

R. Shindel, Senior Reactor Operator

E. Martin, Emergency Diesel Generators System Engineer

A. Bready, Contract Probabilistic Risk Assessment Engineer

M. Azzaro, License Requalification Instructor

B. Tyers, Building Equipment Drains System Engineer

D. Price, Manager Outage Services

M. Crisafulli, Mechanical Maintenance Superintendent

J. Louch, Manager Electrical Maintenance

J. Lewis, Project Manager Reactor

T. Wallender, Project Manager ISFSI

P. Marconni, Dry Cask Storage Loading Supervisor

J. Harris, ALARA Engineer

F. Foster, Operations Maintenance and Technical Instructor

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened NoneOpened/Closed05000354/2006-005-00LER Drywell Hoist Breaker Not Open Prior to Mode 2Entry (Section 4OA3.2)

Closed

05000354/FIN-2006015-03URIInspection of PRA Quality Issues, and NRC Reviewof HEP Assigned Value for Battery Charger Cross-

tie Operator Action (Section 4OA5.2)

A-2Attachment

Discussed

None

LIST OF DOCUMENTS REVIEWED

In addition to the documents identified in the body of this report, the inspectors reviewed thefollowing documents and records:Hope Creek Generating Station (HCGS) Updated Final Safety Analysis ReportTechnical Specification Action Statement Log (SH.OP-AP.ZZ-108)

HCGS Nuclear Controls Operator (NCO) Narrative Logs

HCGS Plant Status Reports Weekly Reactor Engineering Guidance to Hope Creek Operations Hope Creek Operations Night Orders and Temporary Standing Orders

Section 1R01: Adverse Weather ProtectionProceduresSH.OP-DG.ZZ-0011, Rev. 5, Station Seasonal Readiness GuideHC.OP-AB.MISC-0001, Rev. 8, Acts of Nature

HC.OP-GP.ZZ-0003, Rev. 18, Station Preparations for Winter Conditions

NC.OP-DG.ZZ-0002, Rev. 6, Severe Weather GuideNotifications203138002031399120317773

Other DocumentsDE-CB.EA/EP/EQ-0052, Rev. 2, Configuration Baseline Documentation for Station ServiceWaterHope Creek Winter/Grassing Readiness Report and Weekly Update InformationSection 1R04: Equipment AlignmentProceduresHC.OP-AB.ZZ-0172, Rev. 2, Loss of 4.16KV Bus 10A403 C ChannelHC.OP-GP.PB-0003, Rev. 11, 4.16KV Bus 10A403 Removal And Return To Service - C

ChannelHC.OP-AP.ZZ-0108, Rev. 27, Operability Assessment and Equipment Control Program

HC.OP-SO.EA-0001, Rev. 29, Service Water System Operation

HC.OP-SO.EP-0001, Rev. 15, Service Water Traveling Screens System Operation

HC.OP-SO.EG-0001, Rev. 38, Safety & Turbine Auxiliaries Cooling Water System Operation

HC.OP-SO.GJ-0001, Rev 44, Control Area Chilled Water System OperationHC.OP-SO.GU-0001, Rev 23, Filtration, Recirculation and Ventilation System OperationHC.OP-SO.GK-0001, Rev 11, Control Area Ventilation System OperationHC.OP-ST.KJ-0003, Rev. 61, Emergency Diesel Generator 1CG400 Operability Test - Monthly

A-3Attachment Completed SurveillancesHC.OP-ST.KJ-0002, Emergency Diesel Generator 1BG400 Operability Test - Monthly, dated2/16/07HC.OP-ST.KJ-0004, Emergency Diesel Generator 1DG400 Operability Test - Monthly, dated

2/19/07DrawingsM-90-1, Sheet 2, Rev 20, Auxiliary Building Control Area Chilled Water Systems, Control Area Chillers

M-90-1, Sheet 3, Rev 17, Auxiliary Building Control Area Chilled Water System

M-78-1, Sheet 1, Rev 15, Aux. Bldg. Control Area Air Flow DiagramNotifications201746512031409920314131203142292031444220317161203173062031400620315597203156122031589320317545

2031769420317939Orders301350806006820760068187501026965010269750102711

Other DocumentsWCD 4193678

Section 1R05: Fire ProtectionProceduresHope Creek Pre-Fire Plan

FRH-II-412, Rev. 3, RCIC Pump & Turbine Room, RHR Pump & HeatExchanger Rooms & Electrical Equipment Room Elevation: 54'-0" Hope Creek Pre-Fire Plan

FRH-II-571, Rev. 5, HVAC Equipment Rooms Elevation: 178' & 199' Hope Creek Pre-Fire Plan

FRH-II-563, Rev. 6, Control Area HVAC Equipment RoomsElevations: 155'-3" & 175'-0"Hope Creek Pre-Fire Plan

FRH-II-552, Rev. 7, Control Room & Electrical Access Area Elevation:137'-0"Hope Creek Pre-Fire Plan

FRH-III-133, Rev. 6, Turbine Building Elevation: 102'-0"Hope Creek Pre-Fire Plan

FRH-II-531, Rev. 7, Diesel Generator Rooms, Elevation: 102'-0" Hope Creek Pre-Fire Plan

FRH-II-471, Rev. 3, Refuel Floor, Elevation: 201'-0" Hope Creek Pre-Fire Plan

FRH-II-424, Rev. 3, MCC Area, Elevation: 77'-0" Hope Creek Pre-Fire Plan

FRH-II-431, Rev. 3, MCC Area, Elevation: 102'-0" Hope Creek Pre-Fire Plan

FRH-II-151, Rev. 4, Turbine Building, Elevation: 137'-0"

HC.FP-AP.ZZ-0004, Rev. 10, Actions for Inoperable Fire Protection - Hope Creek Station Salem and Hope Creek Fire Impairment Log Book, dated 2/23/07

HC.FP-SV.ZZ-0056, Rev. 3, Fire Barrier Inspection

HC.FP-SV.KC-0066, Rev. 3, Control Room Halon Storage Cylinders Volume Check

HC.FP-ST.KC-0048, Halon System Air Flow Test, dated 4/14/06

HC.FP-SV.KC-0066, Control Room Halon Storage Cylinders Volume Check, dated 5/21/06 &11/15/06

A-4Attachment Notifications203138842031393920314142203144042031454320314545202463332031398520314098203143952031443420314544

246331202463342024633520319478Other DocumentsWCD 4196769

Section 1R06: Flood Protection MeasuresProceduresHC.RW-FT.HB-0001, Rev. 0, Sump Pump Status Check - MonthlyHC.RW-SO.HG-0001, Rev. 5, Radioactive Drains and Waste System Operation

HC.ER-DG.ZZ-0002, Rev. 2, System Function Level Maintenance Rule Scoping vs. RiskReferenceCalculationsCALC. No. 11-92, Rev. 5, Reactor BLDG Flooding -

EL 54' & 77'CALC. No. 11-0067, Rev. 1, High Energy Line Break Analysis in Reactor BuildingCompleted SurveillancesHC.OP-IS.SK-0101, dated 3/9/07, Plant Leak Detection System Valves - Inservice Test DrawingsM-97-1

SH.2, Rev. 15, Building and Equipment Drain Reactor Building Notifications201844692018880320198042202168832023626620263789202814732028415820287943203002102030996620313916Orders70047610700476517005219370062469301087443011970430146717600455256005079460067964Other DocumentsCALC. No. 11-92, Rev. 5, Reactor BLDG Flooding -

EL 54' & 77'CALC. No. 11-0067, Rev. 1, High Energy Line Break Analysis in Reactor Building Completed Surveillance

HC.OP-IS.SK-0101, dated 3/9/07, Plant Leak Detection System Valves

- Inservice Test Hope Creek Generating Station Individual Plant Examination, dated April 1994Operating ExperienceNRC Information Notice 92-69: Water Leakage from Yard Area Through Conduits into Buildings,dated 9/22/92NRC Information Notice 98-31: Fire Protection System Design Deficiencies and Common-ModeFlooding of Emergency Core Cooling System Rooms at Washington Nuclear Project Unit

2, dated 8/18/98

A-5Attachment NRC Information Notice 2005-11: Internal Flooding/Spray-Down of safety-Related EquipmentDue to unsealed Equipment hatch Floor Plugs and/or Blocked Floor Drains, dated 5/6/05

Section 1R11: Licensed Operator Requalification ProgramProceduresSH.OP-AS.ZZ-0001, Rev. 13, Operations StandardsHC.OP-AP.ZZ-0108, Rev. 27, Operability Assessment and Equipment Control Program

HC.OP-AB.ZZ-0000, Rev. 3, Reactor Scram

HC.OP-AB.COOL-0001, Rev. 11, Station Service Water

HC.OP-AB.COOL-0002, Rev. 1, SACS / TACS Cooling

HC.OP-AB.BOP-0002, Rev. 7, Main Turbine

HC.OP-EO.ZZ-0101FC, Rev. 10, Reactor Pressure Vessel (RPV) Control Flow Chart

HC.OP-EO.ZZ-0102FC, Rev. 11, Primary Containment Control Flow ChartNotifications203139152031424920314006

Other DocumentsHCGS Event Classification Guide and ProceduresCompleted Emergency Classification Paperwork

Section 1R12: Maintenance EffectivenessProceduresHC.ER-DG.ZZ-0002, Rev. 2, System Function Level Maintenance Rule Scoping Vs. RiskReference

HC.OP-ST.GS-0003, Rev. 5, Reactor Building/Suppression Chamber Vacuum Breaker Operability Test - MonthlyDrawingsM-57-1, Rev. 25, Containment Atmosphere Control Notifications203078852031391620314363203163522031653220317017203175292031753020317546203176652031772320317954

20318002203180032030923620308540Orders7006494870064928

Other DocumentsContainment Atmosphere Control Maintenance Rule Availability Graphs (October 2005 - January

2007)

IST Component Requirement for 1-GS-HV-5029

IST Component Requirement for 1-BC-HV-F007A

A-6Attachment

Section 1R13: Maintenance Risk Assessments and Emergent Work ControlProceduresHC.OP-AP.ZZ-0108, Rev. 27, Operability And Equipment Control ProgramSH.OP-AP.ZZ-0027, Rev. 12, On-Line Risk Assessment

HC.ER-DG.ZZ-0002, Rev. 2, System Function Level Maintenance Rule Scoping vs. Risk Reference

HC.OP-AB.ZZ-0172, Rev. 2, Loss of 4.16KV Bus 10A403 C Channel

HC.OP-GP.PB-0003, Rev. 11, 4.16KV Bus 10A403 Removal And Return To Service - C

Channel

HC.OP-ST.KJ-0003, Rev. 61, Emergency Diesel Generator 1CG400 Operability Test - Monthly

HC.OP-AB.COMP-0001, Rev. 2, Instrument and/or Service Air

HC.OP-AB.COMP-0002, Rev. 4, Primary Containment Instrument Gas

HC.OP-AB.BOP-0006, Rev. 9, Main Condenser Vacuum

HC.OP-AB.BOP-0003, Rev. 3, Turbine Hydraulic Pressure

HC.OP-SO.CH-0001, Rev. 37, Main Turbine Control Oil (EHC) System OperationCompleted SurveillancesHC.OP-ST.ZZ-0001, dated 3/9-12/07, Power Distribution Lineup- Weekly Notifications2031559720315612203158932031372220313763203144192031795620318063203158912031803120318936Orders501020675010206830101292301350806006820760068187501026965010269750102711301467173014901060068538Other DocumentsHCGS PRA Risk Evaluation Forms for Work Week Nos. 701 - 713HCGS Relay Test Orders

HCGS Relay Work Standards

WCDs

4192439,

4192977, 4193678

Completed Surveillance

HC.OP-ST.ZZ-0001, dated 3/9-12/07, Power Distribution Lineup-

Weekly

Section 1R15: Operability EvaluationsProceduresHC.OP-SO.AF-0001, Rev. 33, Extraction Steam, Heater Vents and Drains System OperationHC.OP-DL.ZZ-0005, Rev. 39, Attachment 9, Operator Action for

FWH Level Outside Normal BandHC.OP-ST-BB-0001, Rev. 35, Recirculation Jet Pump Operability - Daily

HC.OP-AP.ZZ-0108, Rev. 27, Operability Assessment and Equipment Control Program

HC.OP-AB.IC-0001, Rev. 6, Control Rod

A-7Attachment Completed SurveillancesHC.OP-ST.BF-0002, dated 3/27/07, Control Rod Drive Accumulator Operability Check - WeeklyDrawingsM-04-1, Rev.11, Vents & Drains Heaters 3,4,5 & 6E-0018-1, Sh. 1, Rev. 21, Single Line Meter & Relay Diagram 480 Volt Class 1E Unit Substation

10B410, 10B420, 10B430, 10B440, 10B450, 10B460, 10B470, 10B480

M-11-1, Sh. 1, Rev. 29, Safety Auxiliaries Cooling Reactor Building

1-P-EG-06, Sh. 1, Rev. 15, System Isometrics / Reactor Bldg. Safety Auxiliary Cooling System

'A' Pump and Heat Exchangers

FSK-P-1-EA-664-22, Sh. 2, Rev. 22, Small Piping/Intake Structure Lubrication Line From 10T-

544 to Valves

SV-2247B & SV-247D

M-10-1, Sh. 3, Rev. 27, Service WaterNotifications203088142030883920308817203132432031311720293016202982532031080020307589202981602028786020292989

203153642031530620308114203080412031422320314302

2031437820314455203144592031379120314104Orders700418987006498370066093600674907006626130118865700601587005868360068125500892397006455160066868

60067964700665807006661480091785Other DocumentsVTD 10855-M-010, Instructions for Installation, Operation, and Maintenance of ClosedFeedwater HeatersHCGS UFSAR Section 6.8, Filtration, Recirculation, and Ventilating Systems

HCGS UFSAR Section 9.4, Air Conditioning, Heating, Cooling, and Ventilating Systems

VTD

PN1-A41-8010-0042, Rev. 6,

GEK-90333A, Reactor Recirculation System Operating andMaintenance Instructions50.59 Review Form for Rev. 35 to HC.OP-ST-BB-0001

'B' FRVS Charcoal Radioiodine Test Report Dated March 8, 2007DEH070001, Technical Evaluation for Pipe Hanger Restraint Pin Missing in Hope Creek1-P-EA-142-C001, Design Calculation for Service Water, Lube Water Supply Header andSupports 1-P-EA-142-H01 & H02NRC Inspection Manual Chapter 9900 Technical Guidance: Operability Determinations &

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A-10Attachment

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5NRC Generic Letter 90-05: Guidance For Performing Temporary Non-Code repair of ASMECode Class 1, 2,, and 3 Piping, dated 6/15/90Condition Resolution Operability Determination Notebook

SH.MD-GP.ZZ-0240, System pressure Test Data Sheet, dated 2/23/07Section 1R17: Permanent Plant ModificationsProceduresHC.OP-SO.DA-0001, Rev. 37, Circulating Water System OperationDrawingsE-0203-0, Rev. 3, Cooling Tower Basin Miscellaneous Valve & Intake Structure Deicing ValveJ-09-0, sheet 19, Rev. 2, Circulating Water System Intake Structure Deicing ValvesOrders6006556180090515

Other DocumentsCalculation

EA-0020, Deicing Line Hydraulics

Section 1R19: Post-Maintenance TestingProceduresER-AA-430, Rev. 2, Conduct of Flow Accelerated Corrosion ActivitiesER-HC-430-9055, Rev. 0, Hope Creek Conduct of Flow Accelerated Corrosion Activities

ER-AA-430-1001, Rev. 2, Guidelines for Flow Accelerated Corrosion Activities

HC.MD-CM.BB-0008, Rev. 1, Reactor Recirculation Pump N-7500 Mechanical Seal Rebuild

HC.MD-CM.BB-0003, Rev. 17, Reactor Recirculation Pump Seal Changeout

NC.NA-AP.ZZ-0050, Rev. 7, Station Post Maintenance Testing

NC.MD-AP.ZZ-0050, Rev. 9, Maintenance Testing Program Matrix

SH.MD-AP.ZZ-0003, Rev. 17, Maintenance Department Written Instruction Use Standard

HC.IC-GP.ZZ-0031, Rev. 16, Bailey / NLI Logic Module, Type 862

HC.IC-GP.ZZ-0070, Rev. 5, Bailey Fuse Module, Type 862

SH.MD-EU.ZZ-0002, Rev. 1, Coupling Alignment

SH.MD-GP.ZZ-0240, Rev. 7, System Pressure Test at Normal Operating Pressure and TemperatureDrawings

A-11Attachment M-02-1, Rev.13, Extraction SteamM-43-1, Sh. 1, Rev. 31, Reactor Recirculation

M-10-1, Sh. 2, Rev. 36, Service Water

M-15-0, Sh. 5, Rev. 6, Compressed Air (Instrument)Notifications202838862031140620311515203136542030826720307673202857452025129020281533203091612029947720240048

203160082031375020313869203139222031424720314417

20317835Orders700574056006686660063938301396148009155160066957700501176005712970050000700500186006694070056604

500820465008204250082043600669556006717060056618

800834986006577960055819Other DocumentsClearance Order 4021446MPR Associates 3A Feedwater Heater Pipe Tee Temporary Repair Analysis, Dated January 27, 2007Ultrasonic Test Results for 3A and 3C Feedwater Heater Pipe Tees Design Change Package

80083498, Replacement of SSW Emergency Isolation Drain Valves with AOVs

VTD PM0150Q-0050, Rev. 2, Primary Containment Vacuum Relief Valve Instruction Manual

VTD PJ200Q-2385, Rev. 12, 862 - Cabinet Layout BC652-9

VTD PJ200Q-0384, Rev. 6, 862 System Containment Atmosphere Control H2/O2 Analyzer IslnValve

HV-4955BVTD PJ200Q-0392, Rev. 6, 862 System Containment Atmosphere Control H2/O2 Analyzer IslnValve

HV-5019BVTD PJ200Q-0389, Rev. 7, 862 System Containment Atmosphere Control H2/O2 Analyzer IslnValve

HV-4959BVTD PJ200Q-0393, Rev. 8, 862 System Containment Atmosphere Control H2/O2 Analyzer IslnValve

HV-4966BVTD PJ200Q-1479, Rev. 6, 862 System Containment Atmosphere Control H2/O2 Analyzer IslnValve Intlk. Ch. 6

Section 1R20: Refueling and Outage ActivitiesProceduresHC.OP-SO.AE-0001, Rev. 45, Feedwater System OperationHC.OP-IO.ZZ-0002, Rev. 46, Preparation for Plant Startup

HC.OP-IO.ZZ-0003, Rev. 75, Startup from Cold Shutdown to Rated Power

HC.OP-IO.ZZ-0004, Rev. 68, Shutdown from Rated Power to Cold Shutdown

HC.OP-IO.ZZ-0007, Rev. 22, Operations from Hot Standby

HC.OP-IO.ZZ-0010, Rev. 6, Scram Recovery

HU-AA-101, Rev. 3, Human Performance Tools and Verification Practices

2Attachment DrawingsM-05-1, Sh. 2, Rev. 23, CondensateNotifications2031149420311443203115532031174320311390203114222031153620311425203136562028574520308267Orders6006743580091551600669576006686670065696

Other DocumentsShutdown Safety Assessment Report for Planned Outage (F71) scheduled to begin 01/26/07and end 1/30/07

Planned Outage Shutdown and Startup Fuel Defect Sampling Plan

10855-D3.28, Design, Installation and Test Specification of the Nuclear Boiler System Ultrasonic Thickness Examination Records D-58M / D-54B and D-1M / D-1B

Section 1R22: Surveillance TestingProceduresHC.OP-IS.BJ-0001, Rev. 48,

HPCI Main and Booster Pump Set - 0P204 and 0P217 - InserviceTest

HC.OP-ST.KJ-0002, Rev. 61, Emergency Diesel Generator 1BG400 Operability Test - Monthly

HC.OP-ST.KJ-0015, Rev. 24, EDG 1BG400 - 24 Hour Operability Run and Hot Restart Test

HC.OP-ST.BE-0002, Rev. 40, B & D Core Spray Pumps - BP206 and DP206 - In-Service Test

HC.OP-ST.BF-0001, Rev. 23, Control Rod Scram Time Surveillance

HC.MD-ST.PK-0002, Rev 29, 125 Volt Quarterly Battery SurveillanceDrawingsM-55-1, Sh. 1, Rev. 38, High Pressure Coolant InjectionM-56-1, Sh. 1, Rev. 31, HPCI Pump Turbine

E-0009-1, Single Line Meter & Relay Diagram, 125V DC System Channels C & DNotifications203103732030991820296659202862402031372820314020203142522031430520290840202924882029567220299811

203086262029973020292871202924322031608020316111

2031640420293974Orders500997365010147050087561501016746006499550099699

70055909Other DocumentsH-1-BE-NEE-0506, Seismic Evaluation of the Vibration Damper and Absorber Installed on theCore Spray Pumps, 1BP206 and 1DP206

H-1-BE-SDC-0739, Core Spray Pump Absorber Decoupling Evaluation

A-13Attachment

Section 1R23: Temporary Plant ModificationsProceduresNC.CA-DG.ZZ-0103, Rev. 1, Adverse Condition Monitoring and Contingency PlanningDrawingsE-0436-0, Sh. 1, Rev. 7, Electrical Schematic Diagram 4.16KV Class 1E Ckt. Brkr. ControlChiller Compressor Motor 1AK400

E-0436-0, Sh. 2, Rev. 8, Electrical Schematic Diagram 4.16KV Class 1E Ckt. Brkr. Control Chiller Compressor Motor 1AK400Notifications

20315548

Orders8009194660068169

Other DocumentsVTD PM723Q-0013, Sh. 1, Rev. 13, 19FA Electronic Control Diagram for Nuclear Plant DutyVTD PM723Q-0017, Sh. 0, Rev. 5, Logic Control Annunciation Diagram for 19FQ Machine EmergencySection 1EP2: Alert and Notification System TestingOther DocumentsFinal Rep - 10 Design Review ReportSiren Test Results from 2006 & 2007 (bi-weekly silent test & quarterly audible test)

Maintenance Records from November 2005

Section 1EP3: Emergency Response Organization AugmentationOther DocumentsSalem/Hope Creek Emergency PlanERO Member RosterSection 1EP4: Emergency Action Level and Emergency Plan ChangesOther DocumentsAll 50.54(q)

E-Plan and EAL changes from 2005 & 2006

Section 1EP5: Correction of Emergency Preparedness Weaknesses and DeficienciesOrders7005330870053597700539737005404770054318Other Documents

A-14Attachment

LS-AA-120 "Issue Identification and Screening Process," Rev. 6LS-AA-125 "Corrective Action Program (CAP) Procedure," Rev. 11

All Issue Reports related to EP from 12/19/05 - 3/27/07

Drill Critique Reports - 2005 & 2006

50.54(t) Audits done by the Nuclear Oversight Committee (2006 & 2007)

Section 1EP6: Drill EvaluationProceduresSH.OP-AS.ZZ-0001, Rev. 13, Operations StandardsHC.OP-AP.ZZ-0108, Rev. 27, Operability Assessment and Equipment Control Program

HC.OP-AB.ZZ-0000, Rev. 3, Reactor Scram

HC.OP-AB.COOL-0001, Rev. 11, Station Service Water

HC.OP-AB.COOL-0002, Rev. 1, SACS / TACS Cooling

HC.OP-AB.BOP-0002, Rev. 7, Main Turbine

HC.OP-EO.ZZ-0101FC, Rev. 10, Reactor Pressure Vessel (RPV) Control Flow Chart

HC.OP-EO.ZZ-0102FC, Rev. 11, Primary Containment Control Flow ChartNotifications

20314006

Other DocumentsHCGS Event Classification Guide and ProceduresCompleted Emergency Classification Paperwork

Section 2OS1: Access Control to Radiologically Significant AreasProceduresRP-AA-460, Rev. 11, Controls for High and Very High Radiation AreasRP-AA-460,-1001 Rev. 1, Additional High Radiation Exposure ControlsNotifications2030743220306791Section 2OS2:

ALARA Planning and ControlsProceduresRP-AA-270, Rev 3, Prenatal Radiation ExposureRP-AA-220, Rev 3, Bioassay Program

RP-AA-222, Rev 1, Methods for estimating Internal Exposure from In Vivo and In Vitro Bioassay DataRP-AA-400, Rev 4, ALARA Program

RP-AA-401, Rev 7, Operational ALARA Planning and ControlsOther DocumentsALARA Review 2007-25, "A" Rx Recirc Pump Activities

A-15Attachment Micro ALARA Review 2007-27, Drywell EPU Strain Guage RepairSection 2OS3: Radiaiton Monitoring InstrumentationProceduresRP-AA-1001, Rev 0, Establishing Collective radiation Exposure Estimates and Goals

Section 4OA1: Performance Indicator VerificationProceduresLS-AA-2001, Rev. 4, Collecting and Reporting of

NRC Performance Indicator DataLS-AA-2010, Rev. 4, Monthly Data Elements for NRC/WANO Unit/Reactor Shutdown Occurrences

LS-AA-2030, Rev. 4, Monthly Data Elements for NRC Unplanned Power Changes per 7000

Critical Hours

EP-AD-022, " Nuclear Emergency Planning Performance Indicators," Rev. 2Other DocumentsMonthly Operating Reports for the Months of February 2005 through January 2007Hope Creek NRC Performance Indicators, First Quarter 2005 through Fourth Quarter 2006

ERO Drill Participation PI data, 1Q06, 2Q06, 3Q06 & 4Q06

Public Notification System PI data, 1Q06, 2Q06, 3Q06 & 4Q06

DEP PI data, 1Q06, 2Q06, 3Q06 & 4Q06Section 4OA2: Identification and Resolution of ProblemsProceduresHC.IC-AP.ZZ-00017, Rev. 0, Bailey Module Reliability ProgramHC.SE-PR.RL-0001, Rev. 6, Bailey 862 Logic Module Trending ProgramNotifications202992212028215420299528203139692027316520273515202798402027984320279959 202631912024634520239521

2665902026206320308741203085402026702420251290Orders600668606005953960061981600657796006578060065457600611837005456570062185700664638008939880080950

800786548008058370052075700493087004802170052848

8008689070064948600668597005000070050018Other DocumentsRL System Engineer Bailey Solid State Logic Module Failure TrendingVTD PJ200Q-0599, Sh. 0, Rev. 13, 4.16KV System Diesel Generator Circuit Breaker (1)52-

40407

PSEG Electronic System Health Indicator Program (eSHIP)

A-16Attachment

Section 4OA3: Event FollowupProceduresHC.OP-SO.AE-0001, Rev. 45, Feedwater System OperationHC.OP-IO.ZZ-0002, Rev. 46, Preparation for Plant Startup

HC.OP-IO.ZZ-0003, Rev. 75, Startup from Cold Shutdown to Rated Power

HC.OP-IO.ZZ-0004, Rev. 68, Shutdown from Rated Power to Cold Shutdown

HC.OP-IO.ZZ-0007, Rev. 22, Operations from Hot Standby

HC.OP-IO.ZZ-0010, Rev. 6, Scram Recovery

HU-AA-101, Rev. 3, Human Performance Tools and Verification PracticesDrawingsM-05-1, Sh. 1, Rev. 24, CondensateM-05-1, Sh. 2, Rev. 23, Condensate

M-06-1, Sh. 1, Rev. 25, Feedwater

M-16-1, Sh. 1, Rev. 31, Condensate Demineralizer

M-41-1, Sh. 1, Rev. 35, Nuclear BoilerNotifications203136562028574520308267

Orders80091551600669576006686670065696

Other Documents10855-D3.28, Design, Installation and Test Specification of the Nuclear Boiler SystemUltrasonic Thickness Examination Records D-58M / D-54B and D-1M / D-1B

Low Reactor Water Level Scram Root Cause Evaluation Plant Computer System Data Trends

Section 4OA5: Other ActivitiesProceduresNC.MD-PM.DCS-0003, Rev. 3, Sealing, Drying, and Backfilling of a Loaded

MPCNotifications2030716420304882

Orders6006392970063881

Other DocumentsDCS Hose Failure During Blowdown and associated Prompt Investigation ReportContamination Found on

HI-TRAC and associated Apparent Cause Report Dry Cask Storage Lessons Learned Action Items, Canister #1, dated 11/6/2006

Dry Cask Storage Post Job Critique Action Item List, dated 2/7/2007

A-17Attachment ALARA Post Job Reviews, Dry Cask Storage Activities, Casks 1-4Dry Cask Storage Notification SummarySection 4OA7: Licensee-Identified ViolationsProceduresHC.OP-IO.ZZ-0002, Rev 47, Preparation for Plant StartupNotifications

20307894

Orders 70064553

Other DocumentsRoot Cause Investigation Report, 10-H-202 Drywell Hoist Breaker Not Open Prior to Mode 2 Entry

HCGS Licensee Event Report 2006-005-00, Dated February 16, 2007

A-18Attachment

LIST OF ACRONYMS

ADAMSAgencywide Documents Access and Management SystemALARAAs Low As Is Reasonably Achievable

ANSAlert and Notification System

ANSI/ANSAmerican National Standards Institute / American Nuclear Society

CAPCorrective Action Program

CEDECommitted Effective Dose Equivalent

CFRCode of Federal Regulations

CRControl Room

DCSDry Cask Storage

DEPDrill and Exercise Performance

EALEmergency Action Level

ECCSEmergency Core Cooling Systems

EDGEmergency Diesel Generator

EHCElectro-Hydraulic Control

EOFE mergency Operations Facility

EP Emergency Preparedness

EROEmergency Response Organization

FRVSFiltration, Recirculation, and Ventilation System

HCGSHope Creek Generating Station

HEPHuman Error Probability

HPCIHigh Pressure Coolant Injection

HVACHeating, Ventilation and Air Conditioning

IMCInspection Manual Chapter

ISFSIIndependent Spent Fuel Storage Installation

LERLicensee Event Report

LOOPLoss of Offsite Power

MCCMotor Control Center

MPCMulti Purpose Canister

MRMaintenance Rule

MSPIMitigating System Performance Index

NCONuclear Controls Operator

NCVNon-cited Violation

NEINuclear Energy Institute

NRCNuclear Regulatory Commission

OAOther Activities

PARSPublicly Available Records

PIsPerformance Indicators

PRAProbabilistic Risk Assessment

PSEGP ublic Service Enterprise Group Nuclear

LLC [[]]

RACSReactor Auxiliaries Cooling System

RCICReactor Core Isolation Cooling

RHRResidual Heat Removal

RPVReactor Pressure Vessel

SACSSafety Auxiliaries Cooling System

SRASenior Reactor Analyst

A-19Attachment SRVSafety Relief ValveSSCsStructures, Systems, and Components

SSWStation Service Water

TACSTurbine Auxiliaries Cooling System

UFSARUpdated Final Safety Analysis Report

URIUnresolved Item

VHRAVery High Radiation Areas

WCDW ork Clearance Document