IR 05000277/2007002

IR 05000277-07-002 and 05000278-07-002, on 01/01/2007 to 03/31/2007; Peach Bottom Atomic Power Station Units 2 and 3; Licensed Operator Requalification Program, Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems, and E

ML071350471
Person / Time
Site:	Peach Bottom
Issue date:	05/15/2007
From:	Krohn P G Reactor Projects Region 1 Branch 4
To:	Crane C M Exelon Generation Co, Exelon Nuclear
KROHN P G, RI/DRP/PB4/610-337-5120
References
FOIA/PA-2010-0209 IR-07-002
Download: ML071350471 (51)

Text

May 15, 2007

Mr. Christopher M. CranePresident and CNO Exelon Nuclear Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348

SUBJECT: PEACH BOTTOM ATOMIC POWER STATION - NRC INTEGRATEDINSPECTION REPORT 05000277/2007002 AND 05000278/2007002

Dear Mr. Crane:

On March 31, 2007, the United States Nuclear Regulatory Commission (NRC) completed aninspection at your Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3. The enclosed integrated inspection report documents the inspection results, which were discussed on April 20, 2007, with Mr. J. Grimes 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. The report documents two NRC-identified findings and one self-revealing finding of very lowsafety significance (Green). These findings were determined to involve violations of NRC requirements. Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program (CAP), the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, 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, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at Peach Bottom.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 NRC Public Document Room or from the Publicly Available Records (PARS) component of the C. M. Crane2NRC's document system (ADAMS). ADAMS is accessible from the NRC Website athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Paul G. Krohn, ChiefReactor Projects Branch 4 Division of Reactor ProjectsDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56

Enclosures:

Inspection Report 05000277/2007002 and 05000278/2007002

w/Attachment:

Supplemental Information C. M. Crane3cc w/encl:Chief Operating Officer, Exelon Generation Company, LLCSite Vice President, Peach Bottom Atomic Power StationPlant Manager, Peach Bottom Atomic Power StationRegulatory Assurance Manager - Peach BottomAssociate General Counsel, Exelon Generation CompanyManager, Financial Control & Co-Owner AffairsVice President, Licensing and Regulatory AffairsSenior Vice President, Mid-AtlanticSenior Vice President - Operations SupportSenior Vice President, Nuclear ServicesDirector, Licensing and Regulatory AffairsJ. Bradley Fewell, Assistant General Counsel, Exelon Nuclear Manager Licensing, PBAPSDirector, TrainingCorrespondence Control DeskDirector, Bureau of Radiation Protection, Department of Environmental Protection R. McLean, Power Plant and Environmental Review Division (MD)G. Aburn, Maryland Department of EnvironmentT. Snyder, Director, Air and Radiation Management Administration, Maryland Department of the Environment (SLO, MD)Public Service Commission of Maryland, Engineering DivisionBoard of Supervisors, Peach Bottom TownshipB. Ruth, Council Administrator of Harford County CouncilMr. & Mrs. Dennis Hiebert, Peach Bottom AllianceTMI - Alert (TMIA)J. Johnsrud, National Energy Committee, Sierra ClubMr. & Mrs. Kip AdamsE. Epstein, TMI AlertR. Fletcher, Department of Environment, Radiological Health Program C. NRC's document system (ADAMS). ADAMS is accessible from the NRC Website athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Paul G. Krohn, ChiefReactor Projects Branch 4Division of Reactor ProjectsDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56

Enclosures:

Inspection Report 05000277/2007002 and 05000278/2007002

w/Attachment:

Supplemental InformationDistribution w/encl

S. Collins, RAM. Dapas, DRA P. Krohn, DRPR. Fuhrmeister, DRPF. Bower, DRP - NRC Senior Resident InspectorM. Brown - NRC Resident InspectorS. Schmitt, DRP - NRC Resident OAJ. Lamb, RI OEDO J. Hughley, PM, NRRJ. Lubinski, NRRM. Kowal, NRRR. Ennis, NRR H. Chernoff, NRRRegion I Docket Room (with concurrences)ROPreports@nrc.gov (All IRs)SUNSI Review Complete: ___PGK___ (Reviewer's Initials)DOCUMENT NAME: C:\FileNet\ML071350471.wpdAfter declaring this document "An Official Agency Record" it will be released to the Public.To receive a copy of this document, indicate in the box:

" C" = Copy without attachment/enclosure " E" = Copy with attachment/enclosure " N" = No copyOFFICERI/DRP RI/DRPRl/DRP NAMEFBower/PGK forRfuhrmeister/PGK forPKrohn/PGKDATE 05/14/07 05/11/07 05/14/07OFFICIAL RECORD COPYML071350471 EnclosureU. S. NUCLEAR REGULATORY COMMISSIONREGION IDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56Report No.:05000277/2007002 and 05000278/2007002Licensee:Exelon Generation Company, LLCFacility:Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3Location:Delta, PennsylvaniaDates:January 1, 2007 through March 31, 2007Inspectors:F. Bower, Senior Resident InspectorM. Brown, Resident InspectorJ. Caruso, Senior Operations EngineerJ. D'Antonio, Senior Operations EngineerL. Cheung, Reactor InspectorJ. Krafty, DRS, Reactor InspectorK. Mangan, DRS, Senior Reactor InspectorR. Nimitz, Senior Health PhysicistD. Tifft, DRS, Reactor InspectorApproved by:Paul G. Krohn, ChiefReactor Projects Branch 4Division of Reactor Projects Enclosure ii

SUMMARY OF FINDINGS

IR 05000277/2007-002, 05000278/2007-002; 01/01/2007 - 03/31/2007; Peach Bottom AtomicPower Station (PBAPS), Units 2 and 3; Licensed Operator Requalification Program,Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems, and EventFollowup. The report covered a 3-month period of inspection by resident inspectors and announcedinspections by a senior health physicist and six regional specialist inspectors. Three Greenfindings, all of which were NCVs, were identified. The significance of most findings is indicatedby their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,"Significance Determination Process" (SDP). Findings for which the SDP does not apply maybe Green or be assigned a severity level after NRC management review. The NRC's programfor overseeing the safe operation of commercial nuclear power reactors is described inNUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006. A.

Cornerstone: Mitigating Systems and Barrier Integrity

Green.

The inspectors identified a non-cited violation (NCV) of Title 10 of theCode of Federal Regulations (CFR) Part 55.53(e), "Conditions of Licenses,"because Peach Bottom Atomic Power Station (PBAPS) incorrectly creditedindividuals with actively performing the functions of a senior operator (SO) whilethose individuals staffed a position that was not specified in PBAPS's TechnicalSpecifications (TS). Specifically, PBAPS incorrectly credited individuals with performing the functions of a SO while those individuals staffed the workexecution control supervisor (WECS) position. The WECS position is notrequired by PBAPS's TS. Corrective actions included issuing a cease and desistorder to licensed operators to stop crediting time in the WECS position as activetime for maintaining licenses. The finding is more than minor because it impacted the human performanceattribute of the Mitigating Systems cornerstone. In addition, the finding is morethan minor because if left uncorrected, it would become a more safety significantsafety concern. Specifically, although the WECS performs activities important tosafety, the active time credited is not in a position defined by TS that involveddirecting the licensed activities of licensed operators. This finding is related tooperator license conditions and was determined to be of very low safetysignificance (Green) because more than 20 percent of the records reviewed haddeficiencies. (Section 1R11.1)*Green. A self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion V,"Instructions, Procedures, and Drawings," was identified for inadequatesurveillance procedure development that changed the use of a common highpressure coolant injection (HPCI)/reactor core isolation cooling (RCIC) line to thetorus from its original design purpose as a partial-flow flush line, to a full-flow test ivline. The cracked piping to the torus was replaced and this issue was placed intothe corrective action program (CAP) for resolution.This finding is more than minor because it is associated with the design controlattribute of the Barrier Integrity Cornerstone and it affected the objective toprovide reasonable assurance that physical design barriers (primarycontainment) protect the public from radionuclide releases caused by accidentsor events. The Significance Determination Process (SDP) Phase 1 screeningidentified that a Phase 2 analysis was needed because the finding affected twocornerstones, specifically the Mitigating Systems and Barrier Integritycornerstones. However, the senior reactor analysts (SRAs) conducted aPhase 3 evaluation because the issue was too complex to evaluate using thePlant Specific Phase 2 Notebook. For events (large or medium break loss-of-coolant accidents) with the greatest potential consequence, the SRAsdetermined that the probability of a large early release remained very lowbecause existing emergency operating procedures direct reactor operators tomaintain torus level and prevent an increase in core damage frequency byinjecting high pressure service water (HPSW) through the residual heat removal(RHR) system. The Phase 3 SDP evaluation concluded that this finding was ofvery low safety significance (Green). (Section 4OA3.2)

Cornerstone: Public Radiation Safety

Green.

The inspectors identified a NCV of TS 5.4.1.C because procedures foreffluent monitoring were inadequately established and maintained. Specifically,the Quality Assurance required procedures for effluent monitoring wereinadequate to detect non-representative sampling of the 'B' train of the mainstack particulate effluents sampling system. This issue was placed in the CAPfor resolution. This finding is more than minor because it affected the Public Radiation SafetyCornerstone objective to ensure adequate protection of public health and safety. This finding was determined to be of very low safety significance because: 1) itwas not a radioactive material control issue; 2) it did involve the effluent releaseprogram; 3) there was an impaired ability to assess dose; and 4) public radiationdoses did not exceed 10 CFR Part 50, Appendix I values. This finding has across-cutting aspect in the human performance area, resources componentbecause the procedures and training of personnel were inadequate to detect thesample bypass. (Section 2PS1)B.Licensee-Identified Violation A violation of very low safety significance, that was identified by the licensee, has beenreviewed by the inspectors. Corrective actions taken or planned by the licensee havebeen entered into the licensee's CAP. The violation and corrective actions are listed inSection 4OA7 of this report.

Enclosure

REPORT DETAILS

Summary of Plant StatusUnit 2 began the inspection period at 100 percent full rated thermal power (RTP). OnFebruary 16, 2007, power was reduced to approximately 58 percent for maintenance on the 2'B' reactor feed pump, control rod timing, water box cleaning, and other planned maintenanceand testing. The unit was returned to full power on February 17, 2007, where it remainedexcept for brief periods to support planned testing and rod pattern adjustments. OnFebruary 28, 2007, an unplanned power reduction to approximately 76 percent was performedto maintain main condenser vacuum when the 2 'C' circulating water pump tripped. Later onFebruary 28, 2007, the unit returned to full power where it remained until the end of theinspection period.Unit 3 began the period at 100 percent RTP. On January 12, 2007, power was reduced toapproximately 58 percent for maintenance on the 3 'C' reactor feed pump, control rod timing,and other planned maintenance and testing. The unit returned to full power on January 13,2007, where it remained except for brief periods to support planned testing and rod patternadjustments. On February 27, 2007, an unusual event (UE) was declared in response to a firein non-safety-related switchgear located in the turbine building. Consequently, an unplannedpower reduction to approximately 55 percent was performed due to the fire-induced loss ofisophase bus duct cooling. Subsequently, power was further reduced to 50 percent followingan unplanned trip of the 3 'B' reactor feed pump. On February 28, 2007, power was increasedto 90 percent following the return of isophase bus duct cooling. On March 2, 2007, the unit wasreturned to full power where it remained until the end of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 Sample)

a. Inspection Scope

The inspectors reviewed one sample of PBAPS's preparation for frazzle ice conditions. The inspectors reviewed abnormal operations procedure (AO)-29.2, "Discharge Canal toIntake Pond Cross-Tie Gate Operation and Frazzle Ice Mitigation," Revision 12, toensure PBAPS appropriately prepared for environmental conditions conducive to theformation of frazzle ice. The inspectors discussed PBAPS's actions with maintenanceand engineering personnel. Documents reviewed during this inspection are listed in theAttachment.

b. Findings

No findings of significance were identified.

2Enclosure1R02Evaluations of Changes, Tests, or Experiments (71111.02 - 20 Samples: 4 Safety Evaluations; 16 Screening Evaluations)

a. Inspection Scope

The inspectors reviewed four safety evaluations (SEs) completed during the past twoyears. The SEs reviewed were in the Initiating Events and Mitigating Systemscornerstones. The selected SEs were reviewed to verify that changes to the facility orprocedures as described in the Updated Final Safety Analysis Reports (UFSAR) werereviewed and documented in accordance with 10 CFR Part 50.59, and that the safetyissues pertinent to the changes were properly resolved or adequately addressed. Thereviews included the verification that PBAPS had appropriately concluded that thechanges could be accomplished without obtaining license amendments. The inspectors also reviewed 16 screening evaluations for changes, tests andexperiments for which PBAPS determined that SEs were not required. This review wasperformed to verify that the threshold for performing SEs was consistent with 10 CFR Part 50.59. The documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04Q - 4 Partial Walkdown Samples)Partial Walkdown

a. Inspection Scope

The inspectors performed a partial walkdown of four systems to verify the operability ofredundant or diverse trains and components when safety-related equipment wasinoperable. The inspectors performed walkdowns to identify any discrepancies thatcould impact the function of the system and potentially increase risk. The inspectorsreviewed applicable operating procedures, walked down system components, andverified that selected breakers, valves, and support equipment were in the correctposition to support system operation. The inspectors also verified that PBAPS hadproperly identified and resolved equipment alignment problems that could causeinitiating events or impact the capability of mitigating systems or barriers and enteredthem into the CAP. The four systems reviewed were:*Unit 3 'B' Core Spray Pump with the 3 'A' Core Spray Pump Out-of-Service;*'B' Emergency Service Water (ESW) Pump with the 'A' ESW PumpOut-of-Service for Breaker Maintenance;*Unit 2 'A' RHR Loop With the Unit 2 'B' RHR Loop Out-of-Service; and*Standby Gas Treatment System with secondary containment breached for Unit 2 and Unit 3.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - 9 Samples).1Fire Protection - Tours

a. Inspection Scope

The inspectors reviewed PBAPS's Fire Protection Plan, Technical Requirements Manual(TRM), and the respective pre-fire action plan procedures to determine the required fireprotection design features, fire area boundaries, and combustible loading requirementsfor the areas examined during this inspection. The fire risk analysis was reviewed togain risk insights regarding the areas selected for inspection. The inspectors performedwalkdowns of nine areas to assess the material condition of active and passive fireprotection systems and features. The inspection was also performed to verify theadequacy of the control of transient combustible material and ignition sources, thecondition of manual firefighting equipment, fire barriers, and the status of any relatedcompensatory measures. The following nine fire areas were reviewed for impaired fireprotection features:*Unit 3 Service Water Screen Wash Pump (Fire Zone 144);*Radwaste Building, Elevations 150' & 165' (Fire Zone 72J);*Unit 2 Reactor Recirculation Pump Motor Generator Set Room (Fire Zone 4C);*Emergency Cooling Tower (Fire Zone 136);*Unit 2 HPCI Pump Room (Fire Zone 59);*Unit 2 'A' & 'C' RHR Pump and heat exchanger (HX) Room (Fire Zone PF-1);*Unit 2 'A' & 'C' Core Spray Rooms (Fire Zone PF-5A);*Unit 3 'A' & 'C' RHR Pump and HX Rooms (PF-11); and*Unit 3 Reactor Building Closed Cooling Water Room (Fire Zone PF-12B).

b. Findings

No findings of significance were identified..2Fire Protection - Drill Observation (71111.05A - 1 Sample)

a. Inspection Scope

The inspectors observed a Unit 3 HPCI pump room fire drill on January 10, 2007. Thedrill simulated a Class B fire (lubricating oil) at the bearings of the Unit 3 HPCI pump dueto a bearing failure. The inspectors evaluated the fire brigade performance during thedrill to assess the readiness of station personnel to fight fires. Specifically, theinspectors verified that:

4Enclosure*The fire brigade (FB) leader responded to the fire area to begin assessing thesimulated fire and establishing a command post;*Security radiation protection personnel and a licensed senior reactoroperator (SRO) (floor supervisor) responded and were available to support theFB leader;*The four FB members donned the applicable turnout gear and responded to thefire area;*Self-contained breathing apparatuses were available and properly worn by thefour FB members;*FB leader maintained command and control of the fire brigade and had a copy ofthe pre-fire plan; *The fire hoses were capable of reaching the fire hazard and were laidappropriately;*The FB used the "two person rule" for personnel safety;*The FB brought sufficient fire fighting equipment to the scene;*Drill personnel followed the scenario and all drill objectives were met; and*The FB and the evaluators performed a post-drill critique and validated that thedrill objectives were met.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06 - 2 Internal Samples)Internal Flooding

a. Inspection Scope

The inspectors reviewed PBAPS's internal flooding analysis contained in the IndividualPlant Examination (IPE) for the Unit 2 and Unit 3 'A' and 'C' RHR pump rooms. Theinspectors also reviewed Design Basis Document (DBD) P-T-09, Revision 8, "InternalHazards." The inspectors walked down Unit 2 and Unit 3 RHR pump rooms to verifyinternal flooding design features were as described in the IPE. The inspectors alsoinspected floor plugs to verify that they were installed in the Unit 2 and Unit 3 'A' and 'C'RHR pump room drains to prevent multiple RHR pumps from being affected by a flood.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07 - 1 Sample)

a. Inspection Scope

Based on a plant specific risk assessment and past inspection results, the inspectorsselected the following heat exchanger for review:

5Enclosure*RT-O-010-660-2, RHR HX Performance Test, Revision 7, completed March 10, 2007.The inspectors reviewed one sample of safety-related HX testing to identify anydegraded performance or potential for common cause problems that could increaseplant risk. The inspectors reviewed the results of testing performed in accordance withPBAPS's procedures. The inspectors reviewed test results and compared them withacceptance criteria contained within the procedure to verify that all acceptance criteriahad been satisfied. The inspectors also reviewed the UFSAR to ensure that HXinspection results were consistent with the design basis.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11B - 1 Sample)

.1 Biennial Review of Licensed Operator Requalification Program

a. Inspection Scope

The inspectors reviewed documentation of operating history since the last requalificationprogram inspection. The inspectors also discussed facility operating events with theresident staff. Documents reviewed included NRC inspection reports, plantperformance insights, licensee event reports (LERs), and licensee issue reports (IRs)that involved human performance issues for licensed operators to ensure thatoperational events were not indicative of possible training deficiencies.The inspectors reviewed three examination sets (weeks 1, 2, and 3) for both thecomprehensive RO and SRO biennial written examinations administered in 2006, aswell as scenarios and job performance measures (JPMs) administered during thiscurrent examination cycle to ensure the quality of these examinations met or exceededthe criteria established in the Examination Standards and 10 CFR Part 55.59. Duringthe onsite weeks of this inspection, the inspectors observed the administration ofoperating examinations to operating crews (PS-1 and 2). The operating examinationsconsisted of two or three simulator scenarios for each crew and one set of five JPMsadministered to each individual. For the site specific simulator, the inspectors observed simulator performance during theconduct of the examinations, and discrepancy reports to verify compliance with therequirements of 10 CFR Part 55.46. The inspectors reviewed simulator maintenance,testing, and control procedures. Simulator maintenance, testing, configuration control,and machine operation were discussed with members of the simulator maintenancestaff. A sample of simulator tests including transients, normal, steady state, andmalfunction tests as well as plant event data comparison tests, were reviewed.

6EnclosureConformance with operator license conditions was verified by reviewing the followingrecords:*Remediation training records for two individual operating examination failures;*Simulator and classroom training attendance records for two training cycles;*Six licensed operator medical records;*Proficiency watch-standing and reactivation records; and *A sample of licensed operator reactivation records.The inspectors interviewed Instructors, training/operations management personnel, andtwo operators for feedback regarding the implementation of the licensed operatorrequalification program to ensure the requalification program was meeting their needsand responsive to their noted deficiencies/recommended changes.The inspectors reviewed a potential examination compromise issue that Exelon self-identified based on a review of recent licensed operator requalification programoperating experience. This item was entered into PBAPS's CAP (IR 545351). On April 13, 2007, the inspectors conducted an in-office review of PBAPS'srequalification examination results. These results included the annual operating testsadministered in 2007. The inspection assessed whether pass rates were consistent withthe guidance of NRC IMC 0609, Appendix I. The inspectors verified that: *Crew failure rate on the dynamic simulator was less than 20 percent. (Failure rate was 0.0 percent);*Individual failure rate on the dynamic simulator test was less than or equal to 20 percent. (Failure rate was 0.0 percent);*Individual failure rate on the walkthrough test (JPMs) was less than or equal to 20 percent. (Failure rate was 0.0 percent);*Individual failure rate on the comprehensive biennial written examination wasless than or equal to 20 percent. (N/A - biennial written examinations were notadministered this examination cycle); and*More than 75 percent of the individuals passed all portions of the examination(100.0 percent of the individuals passed all portions of the examination).The inspectors used the following references as acceptance criteria for the inspection:

• NUREG-1021, "Operator Licensing Examination Standards for Power Reactors,"Revision 9;*Inspection Procedure Attachment 71111.11, "Licensed Operator RequalificationProgram;"*NRC Inspection Manual Chapter (IMC) 0609, Appendix I, "OperatorRequalification Human Performance SDP;" and *10 CFR Part 55.46, "Simulation Facilities."

b. Findings and Observations

Introduction:

The inspectors identified a non-cited violation (NCV) of Title 10 of theCode of Federal Regulations (CFR) Part 55.53(e), "Conditions of Licenses," becausePeach Bottom Atomic Power Station (PBAPS) incorrectly credited individuals withactively performing the functions of a senior operator (SO) while those individualsstaffed a position that was not specified in PBAPS's Technical Specifications (TS). Specifically, PBAPS incorrectly credited individuals with performing the functions of aSO while those individuals staffed the work execution control supervisor (WECS)position.

Description:

During discussions with licensed senior reactor operators (SROs), theinspectors discovered that the SROs were taking credit for maintaining their licenseactive while standing the WECS position. The inspectors determined that ExelonProcedure OP-AA-105-102, "NRC Active License Maintenance," Revision 8, Section4.1.1.1, states that, "The WECS position may also be used to satisfy active licenserequirements, provided at least one shift each quarter is performed in the unit supervisorposition." A review of OP-AA-105-102, "NRC Active License Maintenance," Attachment1, "Active License Tracking Log," found numerous SROs that were incorrectly takingcredit for standing the WECS position; a position that is not required to be licensed perPBAPS's TS. The inspectors reviewed PBAPS's TS and determined from section 5.3.2that PBAPS has only committed to have the minimum on-site staffing required by 10CFR Part 50.54(m).

For a two unit facility with one control room, 10 CFR Part 50.54(m) requires a minimumof two SROs. 10 CFR Part 50.54(m)(ii) requires that one of the SROs be assignedresponsibility for overall plant operation. At PBAPS, that position is held by the shiftmanager. 10 CFR Part 50.54(m)(iii) requires that a person holding a SO license be inthe control room at all times. At PBAPS, that position is held by the unit supervisor(previously the control room supervisor position). Therefore, per 10 CFR Part 50.54(m),PBAPS is only required to have a unit supervisor and a shift manager.10 CFR Part 55.53(e) states, in part, that to maintain active status, the licensee shallactively perform the functions of an operator or SO. 10 CFR Part 55.4 defines "activelyperforming the functions of an operator or SO" as an individual that has a position onthe shift crew that requires the individual to be licensed as defined in the facility's TS,and that individual carries out and is responsible for duties covered by that position. AtPBAPS, the only two positions that are required to be licensed per PBAPS's TS are theunit supervisor and the shift manager. Therefore, the only two positions that should becredited with active license time are the unit supervisor and the shift manager.The performance deficiency is that PBAPS incorrectly credited individuals withperforming the functions of a SO while those individuals staffed the work executioncontrol supervisor (WECS) position.

Analysis:

The finding is more than minor because it impacted the human performanceattribute of the Mitigating Systems cornerstone. In addition, the finding is more than 8Enclosureminor because if left uncorrected, it would become a more safety significant safetyconcern. Specifically, although the WECS performs activities important to safety, theactive time credited was not in a position defined by TS that involved directing thelicensed activities of licensed operators. Traditional enforcement does not applybecause there were no actual safety consequences, impacts on the NRC's ability toperform its regulatory function, or will aspects to the violation. The finding wasevaluated using the NRC IMC 0609, Appendix I. The SDP, Appendix I, Block 24 appliessince the issue is related to the licensee's program for maintaining active operatorlicenses and ensuring the medical fitness of its licensed operators. Since the WECSposition is not required to be licensed by the facility's TS, giving SRO credit for activelyperforming the functions of the WECS would impact the licensee's program formaintaining active operator licenses. Since more than 20 percent of the recordsreviewed indicated deficiencies (Block 27), this finding is of very low safety significance(Green).Enforcement: 10 CFR Part 55.53(e), "Conditions of Licenses," requires, in part, that tomaintain an operator license active, the licensee shall actively perform the functions ofan operator or SO on a minimum of seven 8-hour or five 12-hour shifts per calendarquarter. 10 CFR Part 55.4, "Definitions," states, in part, that actively performing thefunctions of an operator or SO means that an individual has a position on the shift crewthat requires the individual to be licensed as defined in the facility's TS and that theindividual is responsible for the duties covered by that position. Contrary to the above,the inspectors identified that prior to January 27, 2007, PBAPS personnel wereimproperly maintaining operator licenses active by incorrectly crediting individuals withactively performing the functions of a SO while manning a position that was not definedin the facility's TS. Specifically, active time was credited for the WECS position and thisposition is not required to be licensed as defined in PBAPS's TS. Corrective actionsincluded PBAPS issuing a cease and desist order to licensed operators to stop creditingtime in the WECS position as active time for maintaining their licenses. Because thisfinding was of very low safety significance and was entered into PBAPS's CAP(IR 00592412), this violation is being treated as an NCV, consistent with section VI.A.1of the NRC Enforcement Policy: NCV 05000277/2007002-01; 05000278/2007002-01,Non-Technical Specifications Position Incorrectly Credited for Active LicenseMaintenance..2Resident Inspector Quarterly Review (71111.11Q - 1 Sample)

a. Inspection Scope

On March 6, 2007, the inspectors observed operators in the plant's simulator duringlicensed operator requalification training to verify that operators' performance wasadequate and that evaluators were identifying and documenting crew performanceissues. The inspectors verified that performance issues were discussed in the crew'spost-scenario critiques. The inspectors also observed the operators' implementation ofoperating procedures. The inspectors discussed the training, simulator scenarios, and 9Enclosurecritiques with the operators, shift supervision, and the training instructors. Theevaluated scenarios observed for this one sample are listed below: *PSEG0731R, Low Torus Level Condition Requires Emergency Blowdown; and*PSEG0715R, Hydraulic Anticipated Transient Without Scram.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - 2 Samples)Routine Maintenance Effectiveness Issues

a. Inspection Scope

The inspectors reviewed two samples of PBAPS's evaluation of degraded conditionsinvolving safety-related structures, systems, and/or components for maintenanceeffectiveness during this inspection period. The inspectors reviewed PBAPS'simplementation of the Maintenance Rule (MR), and verified that the conditionsassociated with the referenced CRs were evaluated against applicable MR functionalfailure criteria as found in licensee scoping documents and procedures. The inspectorsalso discussed these issues with system engineers and MR coordinators to verify thatthey were tracked against each systems' performance criteria and that the systemswere classified in accordance with MR implementation guidance. Documents reviewedduring the inspection are listed in the Attachment. The following conditions werereviewed:* IR 579872, E-1 Emergency Diesel Generator (EDG) Fuel Oil Leaks; and*IR 554132, Replace 3 'D' RHR HX Floating Head Assembly.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 7 Samples)

a. Inspection Scope

The inspectors reviewed PBAPS's planning and risk management actions for plannedand emergent work activities to assess their management of overall plant risk. Theactivities selected were based on plant maintenance schedules and systems thatcontributed to risk. The inspectors reviewed PBAPS's probabilistic safety assessmentrisk evaluation results forms. The inspectors compared the risk assessment results andthe risk management actions to the requirements of 10 CFR Part 50.65(a)(4),Regulatory Guide (RG) 1.182, "Assessing and Managing Risk Before MaintenanceActivities at Nuclear Power Plants," and procedure WC-AA-101, "On-line Work Control 10EnclosureProcess." The inspectors also reviewed selected control room operating logs, walkeddown protected equipment and maintenance locations, and interviewed personnel. These reviews were performed to determine whether PBAPS properly assessed andmanaged plant risk and performed activities in accordance with applicable TS and workcontrol requirements. The following seven planned and emergent work order (WO) andaction request (AR) activities were reviewed:*WO C0219775, Remove Foreign Material (Garlock Gasket Tool) from the Unit 2Generator Brush Rigging;*WO C0219963, Repair Hydrogen Leak on Unit 2 'D' Main Generator HydrogenCooler;*WO C0219318-26 & -29, Remove and Reinstall Hatch Above 3 'D' RHR atReactor Building, 135' Elevation;*WO C0219318-35 & -36, Remove and Reinstall 3 'D' RHR HX Floating Head;*WO C0220444, 4T4 Bus, Inspect, Rework as Required; *WO C0220652, E-3 EDG Inspections Following Overload Event; and*AR A1607626, Unit 2 HPCI Inoperable Due to AO-2-23-042 Failing Closed.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - 6 Samples)

a. Inspection Scope

The inspectors reviewed six issues to assess the technical adequacy of the evaluations,the use and control of compensatory measures, and compliance with the licensing anddesign bases. Associated adverse condition monitoring plans, engineering technicalevaluations, and operational and technical decision making documents were alsoreviewed. The inspectors verified these processes were performed in accordance withthe applicable procedures. The inspectors used TS, TRM, the UFSAR, and associatedDBDs as references during these reviews. The issues reviewed included:*3 'D' RHR HX Leak (IR 514302);*Emergency cooling Tower (ECT) Freezing Issue (AR A1044572);*Lost Part - 2 'C' RHR HX Plug Insertion Tooling Failed, (AR A1546765);*2 'C' RHR HX Leakage to HPSW Greater than Acceptance Criteria,(AR A1604675);*Unit 2 HPCI Inoperable Due to AO-2-23-042 Failing Closed (AR A1607626); and*2 'D' RHR Room Cooler 2DE058 Heat Transfer Test Unsat (IR 608000).

b. Findings

No findings of significance were identified.

11Enclosure1R17Permanent Plant Modifications (71111.17B - 8 Samples)

a. Inspection Scope

The inspectors reviewed eight design changes that were completed within the past twoyears. The review was performed to verify that the design bases, licensing bases, andperformance capability of risk significant structures, systems, and components (SSCs)had not been degraded as a result of the modifications.The inspectors walked down systems to detect possible abnormal installation conditions. The inspectors reviewed the design inputs, assumptions, and design calculations todetermine the design adequacy. For the replacement components, the inspectorsverified material compatibility and seismic qualification. In addition, the inspectorsreviewed the post-modification testing to determine readiness for operations. The10 CFR Part 50.59 screenings and evaluations for the modifications were reviewed toverify that the plant changes were reviewed and documented in accordance with10 CFR Part 50.59.

Finally, the inspectors reviewed the procedures, drawings, DBDs,and UFSAR sections to verify that the documents were appropriately updated. Themodifications reviewed are listed in Attachment 1. The inspectors reviewed IRs associated with 10 CFR Part 50.59 issues and plantmodification issues to ensure that PBAPS was identifying, evaluating, and correctingproblems associated with these areas, and that the planned or completed correctiveactions for the issues were appropriate.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - 7 Samples)

a. Inspection Scope

The inspectors observed selected portions of post-maintenance testing (PMT) activitiesand reviewed completed test records. The inspectors observed whether the tests wereperformed in accordance with the approved procedures and assessed the adequacy ofthe test methodology based on the scope of maintenance work performed. In addition,the inspectors assessed the test acceptance criteria to verify whether the testdemonstrated that the tested components satisfied the applicable design and licensingbases and the TS requirements. The inspectors reviewed the recorded test data toevaluate whether the acceptance criteria were satisfied. The inspectors reviewed sevenPMTs performed in conjunction with the following maintenance activities:*WO C0220132, 2-5A-K003A Replace Relay and Perform PMT;*WO R0810095, E124-P-A (6244) Perform MCU Inspection;*WO R1011869, CHK-O-33-515A; Disassemble Inspect/Rework;*WO C0219318-19 & -23, Perform 3 'D' RHR HX Leak Repairs; 12Enclosure*WO C0219643, 2AP040 Clean/Inspect/Repack Cylinders (2 'A' SLC Pump);*WO C0220652, 0CG012-DR Inspections on the E-3 Diesel Generator Due toIncomplete Procedure Performance During Testing Results in E-3 GeneratorTrip; and*WO C0220288, Recal/Rework/Replace LS-2-23-090 As Required (U2 HPCI Steam Supply Drain Pot Level).

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 7 Samples)

a. Inspection Scope

The inspectors reviewed and observed portions of selected surveillance tests (STs), andcompared test data with established acceptance criteria to verify the systemsdemonstrated the capability of performing the intended safety functions. The inspectorsalso verified that the systems and components maintained operational readiness, metapplicable TS requirements, and were capable of performing the design basis functions. The seven STs reviewed and observed included:*ST-O-020-560-2, Reactor Coolant Leakage Test [Reactor Coolant System (RCS) Leakage Sample];*ST-O-010-301-3, 'A' RHR Loop Pump, Valve, Flow and Unit Cooler Functionaland Inservice Test (IST) [IST Sample];*ST-O-052-701-2, E-1 Diesel Generator 24-hour Endurance Test;*SI3F-13-83-XXCQ, Calibration Check of RCIC Steam Line High Flow InstrumentDPIS 3-13-83;*ST-O-033-300-2, ESW, Valve, Unit Cooler, and ECT Fans Functional IST;*ST-O-052-212-2, E-2 Diesel Generator Slow Start Full Load and IST Test; and*SI3F-23-82-XXC2, Calibration Check of HPCI Flow Instruments FT 3-23-082,FI/FC 3-23-108, E/S 3-23-143, XS 3-23-144 and FS 3-23-078.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications (71111.23 - 2 Samples)

a. Inspection Scope

The inspectors reviewed two temporary modifications to verify that implementation ofthe modifications did not place the plant in an unsafe condition. The review was alsoconducted to verify that the design bases, licensing bases, and performance capabilityof risk significant SSCs had not been degraded as a result of these modifications. The inspectors verified the modified equipment alignment through control room 13Enclosureinstrumentation observations: UFSAR, drawings, procedures, and WO reviews; andplant walkdowns of accessible equipment. The following temporary modifications werereviewed:*TCCP 07-00080, Temporary Power for 30Y023; and*TCCP 07-00081, Temporary Power for 4-T-4-T-C.

b. Findings

No findings of significance were identified.Cornerstone: Emergency Preparedness1EP6Drill Evaluation (71114.06 - 1 Sample)Simulated Training Exercise

a. Inspection Scope

On January 10, 2007, the inspectors observed one emergency plan training exercisethat simulated control of the Emergency Response Organization by the emergencydirector in the technical support center prior to the emergency operations centeraccepting control. The inspection was conducted to assess personnel performance. The training exercise was performed to provide drill and exercise performance (DEP)opportunities for the DEP performance indicator (PI). The review was conducted toidentify any weaknesses and deficiencies in protective action recommendation (PAR)development and simulated notification activities. The inspectors verified that PARdevelopment was performed in accordance with EP-AA-111, "Emergency Classificationand Protective Action Recommendations," and EP-AA-111-F-08, "Limerick/PeachBottom Plant Based PAR." Event classification and notifications were done inaccordance with EP-AA-1007, "Exelon Nuclear Radiological Emergency Plan Annex forPeach Bottom Atomic Power Station." The inspectors verified that training exerciseevaluators captured the results for calculation of the DEP PI. The inspectors alsoverified that weaknesses or deficiencies were captured for the critique of the trainingexercise. The following simulated events were classified during this one trainingexercise:*FG1 - General Emergency, Fission Product Barrier Status; and*MG1 - General Emergency, Loss of Alternating Current Power.

b. Findings

No findings of significance were identified.

14Enclosure2.RADIATION SAFETYCornerstone: Occupational Radiation Safety [OS]2OS1Access Control to Radiologically-Significant Areas (71121.01 - 1 Sample)

a. Inspection Scope

The inspectors reviewed selected activities and associated documentation in the areaslisted below. The criteria used for the evaluation of PBAPS's performance in theseareas was 10 CFR Part 20, TS, and Exelon procedures. The selected areas were:*Plant Walkdowns; *Radiation Work Permit Reviews; and *Jobs in Progress Reviews.The inspectors walked down selected radiological controlled areas and reviewedhousekeeping, material conditions, posting, barricading, and access controls toradiological areas. The inspectors observed and reviewed ongoing work activitiesassociated with packaging of irradiated hardware for disposal.

b. Findings

No findings of significance were identified. Cornerstone: Public Radiation Safety [PS]2PS1Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems(71122.01 - 10 Samples)

a. Inspection Scope

Inspection Planning and In-office InspectionThe inspectors reviewed the 2004 and 2005 Radiological Effluent Release Reports andRadiological Dose Assessment Reports to verify that the program was implemented asdescribed in the Radiological Effluents TS (RETS) and the Offsite Dose CalculationManual (ODCM). The inspectors also reviewed estimated radiological effluentsreleased and projected dose results for 2006. The inspectors reviewed the reports forsignificant changes to the ODCM and to radioactive waste system design and operation. The inspectors determined whether changes to the ODCM were technically justified anddocumented. Technical justifications were reviewed during the onsite inspection.The inspectors evaluated PBAPS's analysis for any additional discharge pathways as aresult of a spill, leak, routine, normal, abnormal, or unexpected liquid discharge orgaseous discharges, which may have developed since the previous inspection. The 15Enclosureinspectors verified that PBAPS had records on sampling locations, type of monitoring,and frequency of sampling to meet 10 CFR Part 20.1501 requirements.The inspectors determined whether modifications made to radioactive waste systemdesign and operation changed the dose consequence to the public. The inspectorsverified that technical reviews and 10 CFR Part 50.59 reviews were performed. Theinspectors determined whether radioactive liquid and gaseous effluent radiation monitorsetpoint calculation methodology changed since completion of the modifications, andthat PBAPS had set and adjusted its radioactive effluent alarm setpoints in accordancewith the methodology and parameters specified within the current ODCM. The inspectors also reviewed PBAPS's actions to resolve any out-of-specificationinter-laboratory cross-check analysis data for the effluent monitoring program and todetermine if remedial action had been taken for the out-of-specification data.The inspectors reviewed the RETS/ODCM to identify the effluent radiation monitoringsystems and applicable flow measurement devices. The inspectors reviewed anyeffluent radiological occurrence performance indicator incidents for onsite follow-up andreviewed PBAPS self-assessments, audits, and event reports that involvedunanticipated offsite releases of radioactive material. The inspectors reviewed theUFSAR description of all radioactive effluent monitoring and radioactive gaseous andliquid processing systems. The inspectors reviewed the RETS/ODCM to identify the programs for identifyingpotential contaminated spills and leakage, and PBAPS's process for control andassessment. The inspectors determined if any licensee procedures and surveillanceactivities address the ability to identify onsite spills and leaks of contaminated fluids.Problem Identification and ResolutionThe inspectors reviewed PBAPS's self-assessments, audits, licensee event reports, andspecial reports related to the radioactive effluent treatment and monitoring programsince the last inspection to determine if identified problems were entered into the CAPfor resolution. The inspectors interviewed staff and reviewed documents to determine iffollow-up activities were being conducted in an effective and timely mannercommensurate with their importance to safety and risk. The inspectors also reviewedself-assessments, audits, and LERs that may have involved unanticipated offsitereleases of radioactive material. For repetitive deficiencies or significant individualdeficiencies in problem identification and resolution, the inspectors determined ifPBAPS's self-assessment activities were identifying and addressing these deficiencies.The inspectors reviewed a selection of corrective action documents since the previousinspection:*NOS Audit NOSA-PEA-03-08, Radiological Environmental Monitoring Program(REMP), ODCM, Non-radiological Effluent Monitoring, October 2003; 16Enclosure*NOS Audit NOSA-PEA-06-04, Chemistry, Radiological Effluent andEnvironmental Monitoring, May 2006; *NOS Audit PEA-05-08, ODCM, REMP, Effluent and Environmental Monitoring; and*IRs: 196314, 253869, 279624, 1499640, 293360, 319434, 339837, 346400,352961, 353353, 35483356601, 386618, 394522, 363933, 394580, 394604,398636, 454242, 467543, 489045, and 569284. The criteria used in this review is contained in 10 CFR Part 20, TS, and stationprocedures.Onsite InspectionThe inspectors walked down components of the gaseous and liquid release systems(e.g., radiation and flow monitors, filters, tanks, and vessels) to observe current systemconfiguration with respect to the description in the UFSAR. The inspectors observedequipment material condition. The inspectors verified that system components were asdescribed in the ODCM and were used for reduction of activity levels in accordance withthe RETS/ODCM. The inspectors observed routine sample collections from the Unit 2 and Unit 3 plantvents and observed analysis of these samples, and samples of particulate and charcoalcartridges from the main stack. The inspectors reviewed use of radioactive gaseouseffluent treatment equipment in accordance with RETS/ODCM requirements, andreviewed use of systems per ODCM guidance. The inspectors reviewed severalradioactive liquid waste release permits, including projected doses to members of the public.The inspectors reviewed records of releases made with out-of-service effluent radiationmonitors, and PBAPS's actions for these releases, to ensure an adequatedefense-in-depth was maintained against an unmonitored, unanticipated release ofradioactive material to the environment. The inspectors determined compensatorysampling and radiological analyses were conducted at the RETS/ODCM requiredfrequency when effluent monitors were declared out-of-service. For unmonitoredreleases, the inspectors determined if PBAPS performed an evaluation of the type andamount of radioactive material that was released, and the associated projected doses tomembers of the public. The inspectors also determined if PBAPS placed information onleaks or spills into its 10 CFR Part 50.75(g) decommissioning file. The inspectors assessed PBAPS's understanding of the location and construction ofunderground pipes and tanks, and storage pools (spent fuel pool) that containradioactive contaminated liquids. The inspectors evaluated if PBAPS may havepotential unmonitored leakage of contaminated fluids to the groundwater as a result ofdegrading material conditions or aging of facilities. The inspectors evaluated PBAPS's capabilities (such as monitoring wells) of detecting spills or leaks and of identifyinggroundwater radiological contamination both onsite and beyond the owner controlledarea. The inspectors reviewed PBAPS's technical bases for its onsite groundwater 17Enclosuremonitoring program. The inspectors discussed with PBAPS its understanding ofgroundwater flow patterns for the site, and in the event of a spill or leak of radioactivematerial, if PBAPS's staff can estimate the pathway of a plume of contaminated fluidboth onsite and beyond the owner controlled area. The inspectors reviewed the PeachBottom Station Hydro-geologic Investigation Report dated September 1, 2006.The inspectors reviewed changes to the ODCM as well as to the liquid or gaseousradioactive waste system design, procedures, or operation since the last inspection. Foreach system modification and each ODCM revision that impacted effluent monitoring orrelease controls, the inspectors reviewed PBAPS's technical justification to determinewhether the changes affected PBAPS's ability to maintain effluents as low as reasonablyachievable (ALARA) and whether changes made to monitoring instrumentation resultedin a non-representative monitoring of effluents. For significant changes to dose values reported in the Radiological Effluent ReleaseReport from the previous report (2004 versus 2005), the inspectors evaluated thefactors which may have resulted in the change. The inspectors evaluated if the changewas influenced by an operational issue (e.g., fuel integrity, extended outage, or majordecontamination efforts).The inspectors reviewed a selection of 2004, 2005, and 2006 monthly, quarterly, andannual dose calculations to ensure that PBAPS properly calculated the offsite dose(both cumulative and projected) from radiological effluent releases and to determine ifany annual TS/ODCM (i.e., Appendix I to 10 CFR Part 50 values) were exceeded and, ifappropriate, issued a PI report if any quarterly values were exceeded. The inspectorsevaluated the source term used by PBAPS to ensure all applicable radionuclidesdischarged, within delectability standards, were included.The inspectors reviewed air cleaning system ST results (standby gas treatment system,control room) to ensure that system operations were within applicable acceptancecriteria specified in the TS. The inspectors reviewed ST results or the methodologyPBAPS used to determine the stack and vent flow rates. The inspectors verified that theflow rates are consistent with RETS/ODCM or FSAR values. The inspectors reviewed records of instrument calibrations performed since the lastinspection for each point of discharge effluent radiation monitor and flow measurementdevice; reviewed completed system modifications; and reviewed the current effluentradiation monitor alarm setpoint value for agreement with RETS/ODCM requirements.The inspectors reviewed calibration records of radiation measurement (i.e., countingroom) instrumentation associated with effluent monitoring and release activities. Theinspectors reviewed quality control records for the radiation measurement instruments,and looked for indications of degraded instrument performance and the correctiveactions taken.The inspectors reviewed the results of the inter-laboratory comparison program to verifythe quality of radioactive effluent sample analyses performed by PBAPS. The 18Enclosureinspectors reviewed PBAPS's quality control evaluation of the inter-laboratorycomparison test and associated corrective actions for any deficiencies identified. Theinspectors also reviewed PBAPS's assessment of any identified bias in the sampleanalysis results and the overall effect on calculated projected doses to members of the public.The inspectors reviewed the results from Exelon's QA audits to determine whetherPBAPS met the requirements of the RETS/ODCM.

b. Findings

Introduction:

An NRC-identified Green non-cited violation of TS 5.4.1, "Procedures,"was identified associated with inadequately establishing, implementing and maintainingwritten procedures for QA of effluent monitoring. Specifically, procedures for QA ofeffluent monitoring were inadequate to detect non-representative sampling of the 'B'train of the main stack particulate effluents sampling system.

Description:

TS, Section 5.4.1.C requires that written procedures for QA of effluentmonitoring be established, implemented, and maintained. PBAPS collects weeklyparticulate samples of its main stack for use in public dose assessment in accordancewith its ODCM. On March 7, 2007, the NRC inspectors identified thatnon-representative samples of main stack 'B' train particulate effluents were collectedfor the week of February 28, 2007. Regulatory Guide (RG) 4.15, "QA for Radiological Monitoring Programs (NormalOperations) - Effluent Streams and Environmental Monitoring," Revision 1, provides theNRC regulatory position on an acceptable QA Program. RG 4.15 identifies the need forQA procedures for continuous sampling systems, including the need for representativesampling. Exelon committed to implement RG 4.15, in accordance with its Station QAProgram, Revision 71.The NRC identified non-representative sampling of the 'B' train particulate sampler forthe week of February 28, 2007. Subsequently, PBAPS reviewed its main stacksampling results and determined that the main stack 'B' particulate effluent sampler trainalso likely exhibited non-representative sampling during the weeks of November 22,2006; December 6, 2006; December 20, 2006; and February 21, 2007. EffectiveAugust 1, 2006, PBAPS had selected the 'B' train effluent measurements sample datafor use in determination of dose to the public. Prior to August 1, 2006, PBAPS relied ona combination of data from both the 'A' and 'B' train effluents sampling systems in thatmaximum values of releases were used. The 'A' channel did not exhibit bypass. The 'A'and 'B' trains each sample the main stack effluent releases and conservative resultswere used. PBAPS conducted preliminary re-evaluation of projected radiation doses tomembers of the public for 2006 and concluded that no doses in excess of 10 CFR 50,Appendix I, had occurred. PBAPS also re-evaluated the year-to-date projected doses tomembers of the public for calendar year 2007. This re-evaluation also did not identifyany projected doses in excess of 10 CFR 50, Appendix I. In addition, to evaluateextent-of-condition, PBAPS evaluated potential sample bypass, and non-representative 19Enclosuresampling, for both the Unit 2 and Unit 3 plant vent stack 'B' train sampling systems. These vents use the same sampling arrangement as the main stack. PBAPS did notidentify sample bypass for these systems or any apparent dose projection issues sincesamples were also collected from both the 'A' and 'B' trains of these systems for reviewand dose assessment. Since August 1, 2006, PBAPS's procedures specified using the'B' train effluent sample analysis results in the assessment of dose to members of thepublic. Failure to implement adequate QA procedures, as specified in TS for effluentmonitoring, is a performance deficiency in that non-representative sampling of effluentsoccurred for the 'B' train radioactive effluents which was reasonably within PBAPS'sability to foresee and correct, and which should have been prevented.

Analysis:

The finding is not subject to traditional enforcement in that the finding did nothave any actual safety consequence, did not have the potential for impacting the NRC'sability to perform its regulatory function, and there were no willful aspects. The finding was greater than minor because failure to implement adequate QA foreffluent monitoring affected the Public Radiation Safety Cornerstone objective to ensureadequate protection of public health and safety. Specifically, the NRC identified, onMarch 7, 2007, that non-representative sampling of main stack particulate effluents hadoccurred for the week beginning February 28, 2007. Using NRC IMC 0609, Appendix D,this finding was determined to be of very low safety significance (Green), in that: 1) itwas not a radioactive material control issue, 2) it did involve the effluent releaseprogram, 3) there was an impaired ability to assess dose, and 4) public radiation dosesdid not exceed 10 CFR 50, Appendix I values.

The inspectors determined that the cause of this finding was related to the resourcesaspect of the human performance cross-cutting area.The above example of failure to establish and implement adequate procedures for QAof effluent monitoring reflects a finding in the cross-cutting area of human performance. Specifically, procedures and training of personnel were not adequate to detect thissample bypass. Exelon placed this issue into its CAP (IR 600686).Enforcement: TS 5.4.1.C requires that procedures for QA of effluent monitoring beestablished, implemented, and maintained. Contrary to this requirement, prior toMarch 7, 2007, the written procedures for QA of effluent monitoring were inadequate todetect non-representative sampling of the 'B' train of the main stack particulate effluentssampling system. Since August 1, 2006, the 'B' train effluent measurements data wereused for public dose assessment. Because this finding was of very low safetysignificance (Green), and PBAPS entered this finding into its CAP (AR 600686600686, thisviolation is being treated as a NCV consistent with Section VI.A of the NRCEnforcement Policy, NUREG-1600: NCV 05000277/2007002-02;05000278/2007002-02, Exelon Did Not Establish and Implement Adequateprocedures for QA of Effluent Monitoring as Required by TS 5.4.1

.

20Enclosure2PS2Radioactive Material Processing and Transportation (71122.02)

a. Inspection Scope

The inspectors observed the packaging and preparation of a Type B shipping cask forshipment (PW-07-003). The inspectors visually inspected the loaded cask inpreparation for shipment. The inspectors selectively reviewed conformance with theapplicable NRC licensed cask Certificate of Compliance (Certificate No. 5805, Revision23).

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIESCornerstones: Initiating Events, Mitigating Systems, and Barrier integrity4OA1Performance Indicator (PI) Verification (71151 - 6 Samples)

a. Inspection Scope

The inspectors reviewed a sample of PBAPS's submittals for the PIs listed below toverify the accuracy of the data reported. The PI definitions and the guidance containedin Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline,"Revision 4, and licensee procedure LS-AA-2001, "Collecting and Reporting of NRCPerformance Indicator Data," were used to verify procedure and reporting requirementswere met. The inspectors reviewed raw PI data collected since October 2006 andcompared graphical representations from the most recent PI report to the raw data toverify the data was included in the report. The inspectors also examined a selectedsample of operators' logs, LERs, CAP records and procedures to verify the PI data wasappropriately captured for inclusion into the PI report and the individual PIs werecorrectly calculated. The inspectors verified that PBAPS initiated an IR (IR 588926) tocorrect a reporting error regarding the unplanned transients PI. The PIs reviewed were:*Unplanned Scrams per 7,000 Critical Hours (Unit 2 and 3);* Scrams with Loss of Normal Heat Removal (Unit 2 and 3); and* Unplanned Power Changes per 7,000 Critical Hours (Unit 2 and 3).

b. Findings

No findings of significance were identified.

21Enclosure4OA2Identification and Resolution of Problems (71152).1Routine Review of Items Entered Into the CAP

a. Inspection Scope

As required by IP 71152, "Identification and Resolution of Problems," and in order tohelp identify repetitive equipment failures, human performance issues or program issuesfor follow-up, the inspectors performed routine screening of issues entered intoPBAPS's CAP. This review was accomplished by selectively reviewing copies of IRs,attending daily screening meetings, and accessing PBAPS's computerized database.

b. Findings

No findings of significance were identified.4OA3Event Followup (71153 - 5 Samples)

.1 (CLOSED) LER 05000277/2006003-00, Elbow Leak on Piping Attached to SuppressionPool Results in Loss of Containment IntegrityOn October 7, 2006, an Unusual Event was declared for Unit 2 due to a loss of primarycontainment.

The loss of primary containment was a result of the discovery of a leak ina 4 inch diameter pipe in a location external to the pipe's penetration of the primarycontainment suppression pool (i.e., torus). The leaking elbow was replaced and thesimilar pipe on Unit 3 was examined. Walkdowns and ultrasonic testing were performedon similar Unit 2 and 3 torus attached piping. These examinations did not identifysimilar concerns. The corrective actions to resolve the underlying causes of this eventwere entered into the CAP (IR 541265). Additional details regarding this event werepreviously documented in NRC Inspection Report 05000277,278/2006-005. Theenforcement aspects of this finding are discussed in Section 4OA3.2 of this report. ThisLER is closed..2(CLOSED) Unresolved Item (URI) 05000277/20060005-02, Loss of PrimaryContainment IntegrityURI 05000277/20060005-02 was opened in NRC Inspection Report 050000277;05000278/2006005, pending the NRC staffs' characterization of this issue following thereview of PBAPS's technical analyses and other documents. The characterization ofthis issue as a finding and its risk significance are discussed below. This URI is closed.

b. Findings

Introduction:

A self-revealing, Green NCV of 10 CFR Part 50, Appendix B, Criterion V,"Instructions, Procedures, and Drawings," was identified for inadequate surveillanceprocedure development that changed the use of a common HPCI/RCIC line to the torusfrom its original design purpose as a partial-flow flush line, to a full-flow test line.

22EnclosureDescription: As previously discussed, on October 7, 2006, PBAPS personneldiscovered a leak in piping attached to the Unit 2 suppression pool that resulted in aloss of primary containment integrity. The leaking piping was the HPCI/RCIC torus flushline. The leak occurred on the intrados of a 45 degree elbow in the 4 inch nominalpiping and was located approximately one foot above the torus penetration (i.e., the leakwas outside of primary containment). The cracks in the elbow resulted from excessivelyhigh flow rates, cavitation, and turbulence. The inspectors reviewed LER 05000277/2006003-00 and PBAPS's root causeinvestigation report (IR 541265-29) to understand the underlying causes for this event. The inspectors noted that the licensee-identified root cause for this self-revealing eventwas inadequate surveillance procedure development and approval that changed the useof this common HPCI/RCIC line to the torus from its original design purpose as apartial-flow flush line, to a full-flow test line. Operation of this piping at flow velocitieshigher than intended was not identified when the ST frequency was increased.

The inspectors noted that the vendor instructions for HPCI system operation andmaintenance were provided to PBAPS in GEK-9682, "Operations and MaintenanceInstructions, High Pressure Coolant Injection System for Peach Bottom Atomic PowerStation, Units 2 and 3," dated February 1971. GEK-9682,Section IV, MaintenanceInstructions, Subsection 4-4, "Flow Test," provides a procedure for full flow testing of theHPCI system. The procedure provides direction to operate the HPCI turbine at reducedspeed (1000-1500 rpm) to limit flow while flushing water to the suppression pool throughboth the minimum flow bypass line and the torus flush line. Subsequently, theprocedure directs isolation of the torus flush line to the suppression pool and opening ofthe test bypass return line to the condensate storage tank before turbine speed isincreased to achieve the full pump flow rate of 5000 gpm. PBAPS's ST procedure, ST-O-023-301-2, "HPCI Pump, Valve, Flow and Unit CoolerFunctional and In-Service Test," steps 6.5.23 to 6.5.26, provided instructions for aligningthe HPCI pump to discharge to the suppression pool at reduced speed and flow throughboth the minimum flow bypass line and the flush line. However, subsequent steps6.5.27 to 6.5.31 did not direct isolation of the torus flush line to the suppression poolbefore turbine speed was increased to achieve full rated pump flow of 5000 gpm. TheST did not limit the flow rate through the flush line to the torus as intended byGEK-9682. The inspectors reviewed a technical evaluation (IR 541265-61) that identified initiatingevents where the existing through-wall cracks in the common HPCI/RCIC line would failand provide a flow path from inside the torus to outside the torus. The evaluationassumed that flow through the drywell to torus downcomers or through the safety reliefvalve (SRV) tailpipes would cause sufficient hydrodynamic load to result in the failure ofthis pipe. The inspectors also reviewed a technical evaluation (IR 541265-62) thatdetermined the amount of time required to lower suppression pool level and uncover thecommon HPCI/RCIC line, assuming no inventory make-up.

23EnclosureThe performance deficiency was inadequate surveillance procedure development andapproval that changed the use of a common HPCI/RCIC line to the torus from itsoriginal design purpose as a partial-flow flush line, to a full-flow test line.

Analysis:

The finding is not subject to traditional enforcement in that the finding did nothave any actual safety consequence, did not have the potential for impacting the NRC'sability to perform its regulatory function, and there were no willful aspects. The finding ismore than minor because it is associated with the design control attribute of the BarrierIntegrity Cornerstone and affected the objective to provide reasonable assurance thatphysical design barriers (primary containment) protect the public from radio nuclidereleases caused by accidents or events.The inspectors evaluated the finding in accordance with IMC 0609, Appendix A,"Significance Determination of Reactor Inspection Findings for At-Power Situations." The SDP Phase 1 screening identified that a Phase 2 analysis was needed because thefinding affected two Cornerstones, specifically the Mitigating Systems cornerstone andthe Barrier Integrity cornerstone. However, the SRA conducted a Phase 3 evaluationbecause the issue was too complex to evaluate using the Plant Specific Phase 2Notebook.Using the site-specific Peach Bottom Standardized Plant Analysis Risk Model, Revision3.21, the SRA made the following assumptions to evaluate this finding:*The exposure time of one-year was used in conducting the evaluation;*A hydrodynamic load (greater than 6 psig) in the torus would occur from a largeor medium break loss-of-coolant accident (LOCA) or a SRV actuation. This loadwould be sufficient to cause torus water level to decrease, uncovering thedowncomer from the drywell and HPCI/RCIC pipe;*Operator action, directed in the emergency operating procedures (EOPs), wouldrecover torus level. If low torus level is indicated in the main control room, then ROs would be directed by the EOPs to maintain torus level using the HPSWsystem through the RHR system and/or to cease injecting to the RCS from thetorus to prevent damaging the injection pumps due to the low level. The failureof operators to perform these actions would cause an increase in CDF andincrease the probability of post vessel breach release from containment (LERF);*For non-LOCA initiating events - if power conversion systems fail or wereassumed to fail due to the initiating event, an SRV would lift. The containmentwould pressurize if suppression pool cooling failed. This would increase theprobability of a containment release (delta LERF) through the pipe break ifcontainment venting was successful (I.e., containment did not fail, prior to coredamage) and torus water level was lower than the pipe at the time of reactorvessel breach. This event does not cause an increase in delta CDF because themitigating systems rely on the condensate storage tank as the primary source ofwater for RCS injection.The SRA developed a HPSW/torus fill fault tree to model the torus pipe failure. Thefault tree included a basic event that would question the tree if only the torus pipe was 24Enclosureassumed to fail and modeled human action and motor operated valves with their electricdependency. The SRA determined that this finding was of very low safety significance (Green),represented a very low change in delta CDF (low to mid 1X10E-8), and a very lowchange of high 1X10E-8 in LERF (delta LERF). The most dominant Phase 3 coredamage sequences involved the initiating events of large and medium LOCAs, and thefailure of the operators to recover torus level. For large and medium LOCA scenarios,the HPSW/torus fill fault tree indicated that success in torus makeup would prevent lossof torus level; however, failing to refill the torus would cause an increase in delta CDFand would result in an increase in delta LERF. For other LERF sequences that did notincrease CDF, the core damage sequences that included SPC failures, successfulcontainment venting (CV), and failure of late injection were identified. These sequenceswere then transferred to the torus fill event tree which included the HPSW/torus fill faulttree and resulted in core damage occurring if the torus pipe retained its integrity (basecase). However, if the pipe was assumed to fail, the event tree would calculate theprobability of a release using the delta CDF and assuming that the release factor of 1.0(for Mark I containment). Accident sequences with suppression pool cooling failure andCV failure were not included in the analysis because the containment was assumed tofail if CV failed, thereby, no benefit would result by refilling the torus. A release wouldoccur if the RCS was breached post-core damage.Enforcement: 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, andDrawings," states, that activities affecting quality shall be prescribed by documentedprocedures and shall be accomplished in accordance with these procedures. Theprocedures shall include appropriate acceptance criteria for determining that importantactivities have been satisfactorily accomplished. Vendor document, GEK-9682,provides a procedure for full-flow testing of the HPCI system. However, this procedureprovides direction to operate the HPCI turbine at reduced speed (1000-1500 rpm) tolimit flow while flushing water to the suppression pool through both the minimum flowbypass line and the torus flush line. Subsequently, the procedure directs isolation of thetorus flush line to the suppression pool and opening of the test bypass return line to thecondensate storage tank before turbine speed is increased to achieve the full pump flowrate of 5000 gpm. Contrary to the above, Exelon procedure ST-O-023-301-2 provided instructions foraligning the HPCI pump to discharge through the torus flush line to the suppression poolat full rated pump flow of 5000 gpm. Specifically, not limiting the flow rate through thetorus flush line to the torus as directed by GEK-9682 resulted in excessively high flowrates and cavitation that led to piping erosion and the resultant through-wall leak in thepartial flow flush line to the torus. Because this finding is of very low safety significanceand has been entered into the CAP (IR 5584677), this violation is being treated as aNCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000277/2007002-03 , Failure to Develop and Implement HPCI SurveillanceTesting in a Manner Consistent with Vendor Specified Test Instructions

.

25Enclosure.3Unit 2 - Fire in 480 Volt Non-Vital Load Center - February 27, 2007

a. Inspection Scope

At approximately 9:16 a.m. on February 27, 2007, a fire was suspected to have startedbased on the receipt of numerous secondary plant alarms in the main control room(MCR) and the report of smoke near the '4T4' 480 Volt load center. The inspectorsresponded to the MCR following a site announcement for the fire brigade to respond toa suspected fire in the Unit 3 turbine building. The inspectors monitored the operators'response to the event and the status of plant equipment. The observations wereprimarily focused on the nuclear safety aspects of the plant's and operators' responses. The inspectors also monitored the response of PBAPS's emergency responseorganization to the declaration of an UE. Subsequent to the fire, the inspectors discussed the fire with operations, engineeringand PBAPS management personnel to gain an understanding of the event and toassess their followup actions. The inspectors reviewed operator logs and operators'actions taken in accordance with licensee procedures. Based on the operators'narrative logs, the fire brigade was dispatched to the Unit 3 turbine building atapproximately 9:20 a.m. Fire personnel investigated and notified the MCR that anactual fire existed at 9:38 a.m. An Unusual Event for a fire not extinguished within15 minutes (emergency action level (EAL) HU6) was declared at 9:41 a.m. All state andlocal government notifications were completed by 9:59 a.m. and the NRC HeadquartersOperations Officer was notified of the event at 10:36 a.m. The fire was considered to beextinguished at approximately 10:32 a.m. At 11:37 a.m., the Unusual Event wasterminated. Prior to the report of the potential fire, Unit 3 was operating at full power. As a result offire and the associated response actions, numerous non-safety-related loads poweredby the '4T4' 480 Volt load center were de-energized. Equipment that was de-energizedincluded: the 'B' isophase bus cooler fan, the 'B' drywell chiller, the 'B' recirculationpump speed controller, the leading edge flow meters and the 'B' reactor feed pump. Plant operators took the required TS actions and responded to the equipment losses byperforming controlled reactor power reductions and stabilized the plant at approximately50 percent of rated power.The inspectors verified that the required reports were made during the event and that nofurther reports are planned. The inspectors also verified that this issue (IR 569889) wasplaced into the CAP. Preliminarily, PBAPS has determined that the fire resulted from anapparent mismatch between the ratings of one breaker and its cubicle in the '4T4' 480volt load center. A root cause investigation was ongoing at the end of the inspectionperiod and will be reviewed by the inspectors during a future inspection period.

b. Findings

At the close of this inspection period, the inspectors were reviewing the event andawaiting the results of the root cause evaluation to understand the potentialperformance deficiencies. This issue is unresolved pending review of PBAPS's causalevaluation and corrective actions by the inspectors to characterize the issue. URI 05000277/2007002-04, Incorrect Size Breaker Resulted in a Fire in the '4T4'480 Volt Load Center

..4Personnel Performance - Missed Procedure Step Resulted in Unplanned Overloading ofthe E-3 EDG

a. Inspection Scope

The inspectors reviewed selected applicable plant records, correction action documentsand approved procedures while evaluating the performance of operations personnel inresponse to non-routine evolutions. The inspectors assessed personnel performance todetermine what occurred and how the operators responded, and to determine if plantpersonnel's response was in accordance with plant procedures and training. Thefollowing non-routine evolution was reviewed:*During the conduct of surveillance testing of the E-3 EDG on March 15, 2007, alicensed operator missed the performance of a required step in a supportingsystem operating procedure. The omission of the procedure step placed the E-3EDG in the isochronous mode while synchronized with offsite power through a4 kilovolt (kV) vital bus. This condition resulted in unexpectedly loading the E-3EDG beyond its 30-minute load rating. The ST and supporting proceduresdirected the synchronization of the E-3 EDG to a selected 4 kV bus to pick upthe bus loads. The procedure subsequently directed opening the offsite powerfeeder breaker to the 4 kV vital bus (the missed step) before placing the EDG inthe isochronous mode. PBAPS placed this issue in the CAP by initiatingIR 604364. Prompt corrective actions included the selected implementation ofadditional peer checking of procedure performance place-keeping. The E-3EDG was inspected for potential damage and tested before being returned to anoperable condition in accordance with TS on March 17, 2007. The causalevaluation of this event was ongoing at the end of the inspection period.

b. Findings

At the close of this inspection period, the inspectors were reviewing the event andawaiting the results of the causal evaluation to understand the potential performancedeficiencies. This issue is unresolved pending review of PBAPS's causal evaluation andcorrective actions by the inspectors to characterize the issue. URI 05000277/2007002-05, Missed Procedure Step Resulted in Unplanned Overloading of the E-3 EDG.

27Enclosure.5(CLOSED) LER 05000277/2006001-00, Main Steam Isolation Valves Exceeded TheirAllowable Leakage LimitsOn September 22, 2006, engineering personnel determined that there were multipleleak rate test failures involving the main steam isolation valves (MSIVs). Thisdetermination was based on local leak rate testing performed during the P2R16Refueling Outage. Four of the eight MSIVs were found to be leaking in excess of theirallowable leakage limits, including both the inboard and the outboard MSIVs for the 'D'main steam line. This condition resulted in a degraded plant safety barrier, a conditionprohibited by TSs and a condition that resulted in multiple trains being inoperable in asafety system. The MSIVs were repaired and returned to an operable status. Theas-left leakage rates were restored below the TS allowable limits. The correctiveactions to resolve the underlying causes of this event are in the CAP (IR 534622) andinclude planned actions to minimize the number of times that the valves are stroked formaintenance and testing in a dry condition to minimize accelerated wear of the internals. This finding is more than minor because it had a credible impact on safety, in that, if the'D' main steam line was required to isolate on a containment isolation signal, thepenetration leakage would be greater than the TS allowable limits. Also, for the 'A' and'C' penetrations, if the redundant valve in the penetration did not close on a containmentisolation signal, containment integrity would not be ensured. The finding affects theBarrier Integrity Cornerstone and was considered to have very low safety significance(Green) using Appendix H of the SDP because the likelihood of an accident leading tocore damage was not affected, the probability of early primary containment failure andtherefore a large early release was small. This licensee-identified finding involved aviolation of TS 3.6.1.3, Primary Containment Isolation Valves. The enforcement aspectsof the violation are discussed in Section

4OA7 of this report.

This LER is closed. 4OA6Meetings, Including ExitExit Meeting SummaryOn April 20, 2007, the resident inspectors presented the inspection results to Mr. J. Grimes and other PBAPS staff, who acknowledged the findings. The inspectorsasked the licensee whether any of the material examined during the inspection shouldbe considered proprietary. No proprietary information was identified. 4OA7Licensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by thelicensee and is a violation of NRC requirements which meet the criteria of Section VI ofthe NRC Enforcement Policy, NUREG-1600, for being dispositioned a NCV. *TS 3.6.1.3 requires that penetration flow paths with one or more MSIVs notwithin MSIV leakage rate limits be isolated within eight hours. Contrary to this,for an indeterminate period during the two-year operating cycle beforeSeptember 18, 2006, four MSIVs were not within MSIV leakage rate limits and 28Enclosurethe penetrations were not isolated within eight hours. This was identified in thelicensee's CAP as IR 534622. This finding is of very low safety significancebecause it does not represent an open pathway in the physical integrity of thereactor containment greater than that assumed in the UFSAR, Chapter 14,"Plant Safety Analysis," for radiological consequences.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Exelon Generation Company personnel

J. Grimes, Site Vice President

M. Massaro, Plant Manager

N. Alexakos, Manager, Engineering-Programs

J. Armstrong, Regulatory Assurance Manager

C. Behrend, Engineering Director

C. Jordan, Chemistry Manager

D. Lewis, Operations Director

G. Stathes, Maintenance Director

S. Taylor, Manager, Radiation Protection

A. Wasong, Training Director

T. VanWyen, Operations Training Manager

B. Artus, Principal Requal Training Instructor

R. Tyler, Simulator Supervisor

W. Pilkey, Physician Assistant

J. Verbillis, Examination Developer

J. Chizever, Mechanical Design Engineering

D. Foss, Sr. Regulatory Engineer

A. Franchitti, Electrical Design Engineering

NRC personnel

Mel Gray, DRP, Branch 4, Branch Chief

J. Caruso, Senior Operations Engineer

J. D'Antonio, Senior Operations Engineer

M. Brown, Resident Inspector

F. Bower, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened05000277/2007002-04URIIncorrect Size Breaker Resulted in a Fire inthe '4T4' 480 Volt Load Center (Section 4OA3.3)05000277/2007002-05URIMissed Procedure Step Resulted inUnplanned Overloading of the E-3 EDG(Section 4OA3.4)

A-2AttachmentOpened and

Closed

05000277, 278/2007002-01NCVNon-Technical Specifications PositionIncorrectly Credited for Active LicenseMaintenance (Section 1R11.1)05000277, 278/2007002-02NCVExelon Did Not Establish and ImplementAdequate Procedures for QA of EffluentMonitoring as Required by TS 5.4.1 (Section 2PS1)

05000277/FIN-2007002-03NCVFailure to Develop and Implement HPCISurveillance Testing in a Manner Consistentwith Vendor Specified Test Instructions(Section 4OA3.2)

Closed

05000277, 278/2007002-01NCVNon-Technical Specifications PositionIncorrectly Credited for Active LicenseMaintenance (Section 1R11.1)05000277, 278/2007002-02NCVExelon Did Not Establish and ImplementAdequate Procedures for QA of EffluentMonitoring as Required by TS 5.4.1 (Section 2PS1)

05000277/FIN-2007002-03NCVFailure to Develop and Implement HPCISurveillance Testing in a Manner Consistentwith Vendor Specified Test Instructions(Section 4OA3.2)

Discussed

None.

LIST OF DOCUMENTS REVIEWED

==Section 1R01: Adverse WeatherIR

568034, Evaluate Cross Tie Gate RemovalIR

584869, Station Critique for Discharge Canal Cross-Tie Gate RemovalAR A1596763, Evaluate Cross Tie Gate RemovalRT-O-28B-800-2, River Temperature and Flow MonitoringM-028-001, Discharge Canal to Intake Pond Gate OperationST-C-095-805-2, Liquid Radwaste Discharge==

A-3Attachment

Section 1R02: Evaluation of Changes, Tests, or Experiments10

CFR 50.59 Safety EvaluationsPB-2004-002-E, Installation and Use of the Reactor Cavity Work Platform (RCWP) DuringOutage, Revision 1PB-2005-01-E, Use of GNF2 Lead Use Fuel Assemblies in PB Unit 3 Cycle 16, Revision 0PB-2005-003-E, Adopt SQUG Methodology for Seismic Qualification of Equipment, Revision 0PB-2006-01-E, Application of TRACG04 for Stability Analysis, Revision 010

CFR 50.59 ScreensPB-2004-022-S, ECR

PB-00119 (U3 MPT and UAT SPR Logic Upgrade), Revision 0PB-2005-007-S, HPCI Turbine Vibration, Revision 0PB-2005-009-S, Core Spray Line Break Detection Setpoint Change, Revision 0PB-2005-027-S, Provide OPRM Clarifications in Tech Spec Bases Section 3.3, Revision 0PB-2005-031-S, Restoration of SBO Test Circuit Due to Duct Bank Damage During BRE #3Rock Anchor Drilling, Revision 0PB-2005-033-S, Revise HPSW System Design Press by

RO-2(3)-801 or 2(3) 789, Revision 0PB-2005-042-S, Install Temperature Monitoring in SRV Pilot Valves, Revision 0PB-2005-046-S, Support Replacement of ESW Valve

HV-3-33-518, Revision 0PB-2005-065-S, PBAPS EDG Keep Warm Modifications, Revision 3PB-2005-067-S, RWM Operability Check, Revision 0PB-2005-078-S, Installation of Restricting Orifices in the HPCI Lube Oil System, Revision 0PB-2006-001-S,

SE-10 Procedure Revision, Revision 0PB-2006-006-S, Procedure Creation AO6F-2-2(3), Revision 0PB-2006-018-S, RCWP Jib Crane, Revision 0PB-2006-029-S, Closing Torque Switch Bypass

MO-2-02-053A, Revision 0PB-2006-055-S, E-1 Diesel Aux Pump Abandonment, Revision 0Calculations86-5049524, Summary Report for Peach Bottom BWR RCWP Framing Design, Revision 2Corrective Action Reports

340404

490304

492097

513278

598300*599323*600094*490319*NRC Identified During Inspection Drawings6280-M-37, Diesel Generator Auxiliary Systems (Lube Oil System), Sheet 3, Revision 40

A-4AttachmentSurveillance ProceduresST-O-62A-210-2, RWM Operability Check, Revision 13MiscellaneousGE Letter, Analysis of Postulated Collision between NF400 Mast 762E974G002 and Low ProfileJib Hoist 124D1815G001, dated 3/4/06GE Letter, Lead Test Assembly Licensing, dated 8/24/81GE-NE-0000-003909767-00, Technical Evaluation to Support Introduction of GNF2 Lead UseAssemblies (LUA) in Peach Bottom Atomic Power Station Unit 3, Revision 0GE-NE-0000-0052-5690-R0, TRACG04 DIVOM 10

CFR 50.59 Evaluation Basis, 4/06NEDC-33144P, GNF2 Lead Use Assembly (LUA) for PBAPS Unit 3, Revision 1NEDE-24011-P-A-15, General Electric Standard Application for Reactor Fuel, 9/05NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodologyfor Reload Applications, 8/96PM L-200-VC-4, Limitorque Valve Operator Engineering Reference Manual, Revision 0PM-1076, Impact of RCWP Jib Crane Failure on Fuel Handling Accident Analysis, Revision 0Supporting Information for 50-59 Evaluation No.

PB-2005-01-EUpdated Final Safety Analysis Report, Peach Bottom Atomic Power Station, Revision 20

Section 1R04: Equipment AlignmentSO 14.1.A-3, Revision 3, Core Spray System Alignment for Automatic or Manual OperationCOL 14.1.A-3B, Revision 9, Core Spray System Loop

BCOL 9A.1.A, Revision 9, Standby Gas Treatment System Automatic OperationP&ID M-362, Sheet 2, Revision 60, Core Spray Cooling SystemProtected Equipment Tracking Sheet, PBAPS Unit 2 & Common, dated January 22, 2007Protected Equipment Tracking Sheet, PBAPS Unit 2 & Common, dated January 31, 2007IR

584836, NOS ID:

Protected Equipment List Discrepancies

Section 1R05: Fire ProtectionOP-AA-201-003, Revision 8, Fire Drill PerformanceRT-F-101-922-2, Revision 3, Fire Drill, completed 1/10/07PF-4C, Revision

5, Prefire Strategy Plan Unit 2 Rx Recirc Pump MG Set Room, Radwaste Building, 135' ElevationPF-72J, Revision

1, Prefire Strategy Plan Radwaste Building, 150' & 165' ElevationPF-136, Prefire Strategy Plan, Emergency Cooling Tower, Fire Zone 136PF-59, Revision 4, Prefire Strategy Plan Unit 2 Reactor Building HPCI Room, 88' ElevationPrefire Strategy Plan U/3 RBCCW Room Radwaste Bldg. 116' Elevation, Fire Zone 12B,Revision 3Prefire Strategy Plan 2 'A' & 2 'C' Core Spray Room, RX Building 91' 6" Elevation, Fire Zones5A & 5B, Revision 1Prefire Strategy Plan 2 'A' & 2 'C' RHR Pump and HX Rooms RB2 -

91' 6" Elevation, Revision 2

A-5AttachmentPrefire Strategy Plan 3 'A' & 3 'C' RHR Pump and HX Rooms RB2 -

91' 6" Elevation, Revision 2

Section 1R06: Flood Protection MeasuresDBD

P-T-09, Revision 8, Internal HazardsIPE Section 3.3.8.2.3, "Reactor Building"

Section 1R07: Heat Sink PerformanceRT-O-010-660-2,

RHR Heat Exchanger Performance Test, Revision 7, completed 3/10/07NRC Generic Letter 89-13, Service Water System Problems affecting safety-related equipment

Section 1R11: Licensed Operator Requalification ProgramPSEG0731R, Low Torus Level Condition Requires Emergency BlowdownPSEG0715R, Hydraulic

ATWSRequalification Program ProceduresHR-AA-07-101, Revision 4, "Licensed Operator Medical Examination"OP-AA-105-101, Revision 10, "Administrative Process for NRC License and Medical Requirements"TQ-AA-106, Revision 8, "Licensed Operator Requal Training Program"TQ-AA-106-304, Revision 7, "Licensed Operator Requal Training Examination Development Job Aid"TQ-AA-106-305, Revision 3, "Licensed Operator Requal Training Examination Administration Job Aid"OP-AA-105-102, Revision 8, "NRC Active License Maintenance"Simulator Baseline Review of Documentation for Transient TestsSTRB 05-3 Exelon Nuclear Simulator Testing Review, 6/9/2005STRB 05-6 Exelon Nuclear Simulator Testing Review, undated

A-6AttachmentSimulator Transient TestsB.1.2.8 Maximum Recirculation Suction Break with Loss of Offsite Power

STPT-RRS20 &MAP02, Revision 3, 10/25/2006.B.1.2.6 Turbine Trip Within Bypass Valve Capacity

STPT-MTA04, Revision 2, 10/20/2006B.1.2.5 STPT - Single Recirc Pump Trip, Revision 3, 10/4/2006B.1.2.1 STPT - Manual Scram, Revision 1, 10/04/2006B.1.2.10 SMPT IPM02 MSIV Closure with Failed Open SRV and No High Pressure ECCS,Revision 1, 10/24/2006Simulator Normal Evolution TestsSNOT NOROP 1 Cold S/D to 100% Power, 12/15/2004SNOT NOROP 4 Scram and Restart to 100% Power, 12/15/2005SNOT NOROP 2 Plant S/D and Cooldown, 12/22/03SNOT NOROP 3 no title (includes reactor startup plus ST surveillance procedures for HPCI,RCIC, RHR, CS), 2/7/2007Simulator Steady State TestsSSPT-Heat Bal Simulator Heat Balance Test, Revision 1, 9/11/2006

Simulator Malfunction TestsSMPT

RHR04 RHR Pump Discharge Line Break, Revision 6, 11/28/2006SMPT

VAC01 480VAC Bus Fault, Revision 5, 11/21/2006SMPT

VAC03 480VAC MCC Fault, Revision 5, 10/10/2006SMPT

RPS05 Automatic Scram Circuit Failure, Revision 3, 11/21/2006SMPT

RRS07A Recirc Pump Shaft Seizure, Revision 6, 2/07/07 Plant Event Data Comparison with SimulatorPDRP 04007 Low Pressure Group 1 Unit 2, 2/24/2005PDRP 04009 Condensate Pump Trip, 12/28/2004Open SWRsSWR# 5654 PMS Digital Displays Do Not Work, 12/15/2003SWR# 6550 MS/OG Numac Rad Monitors Screen Broke on a Total of 3, 7/26/2004SWR# 8014 Core Model IssuesClosed Simulator Work Requests (SWRs)SWR# 9272 Rod position indication is blank after a scramSWR# 9632

AO-8098 and 8099 A & C stroke too fastSWR# 9695

ST-R-002-910-2 step 6.1.8 was unsatSWR# 9381 Problems with E324-O-A, VAC03WSWR# 7412 RCIC operates erratically

A-7AttachmentSWR# 7259 Problems noted with loss of Y-34SWR# 6194 Condenser not working correctlySWR# 7736 'A' Condensate string flow drops after FW heater leak

Section 1R12: Maintenance EffectivenessIR 00579872,

E-1 EDG Fuel Oil LeaksRed/Yellow Maintenance Rule (a)(1) Systems - System 52 - EDG Improvement PlanAR A1424883, General Purpose AR for Misc Evals for System 52 IssuesIR

00207837, PBAPS EDG Action PlanIR

00495141, Exhaust System Bolting Disassembly Results in a Large Percentage of the BoltsBreakingAR A1592701, Examine Lower Support Bolting for RHR HX 3 'D'AR A1591784, Replace 3 'D' RHR Heat Exchanger Floating Head AssemblyAR A1558090, Disassemble, Bubble Test, Repair 3 'D' RHR Heat ExchangerAR A1578288, Increased Leak Rate for 3 'D' RHR Heat ExchangerIR

579005,

RIS-9081 Causing HPSW High Rad AlarmIR

578998,

RIS-9082 Causing HPSW High Rad AlarmIR

583564, Unit 2 'B' Loop HPSW High Rad AlarmIR

606881, 3 'D' Train of RHR Has Exceeded MR (A)(1) Performance Criteria

Section 1R13: Maintenance Risk Assessments and Emergent Work ControlC0219963, 2 'D' E001 Heat Exchanger Leak RepairHU-AA-1211, Pre-job Briefing Checklist for Unit 2 Generator Hydrogen Cooler RepairSA-AA-116-2124, Attachments 2 and 3, Job Hazard Analysis Form for Tightening of Hydrogen FlangeOn-Line Maintenance Approval Form, 3 'D'

RHR Secondary Containment Breach, datedJanuary 23, 2007 Barrier Breach Permit 07-6, Hatch 24, dated January 25, 2007IR

199380-37 & 38, PORC 07-02 Action ItemsGP-16, Breaching and Establishing Secondary Containment, Revision 28Pre-Job Briefing ChecklistHLA/IPA Briefing WorksheetEvaluation of Voluntary Entry into Tech Spec Action Statements for Secondary Containment toSupport RHR Heat Exchanger Corrective Maintenance Work, Revision 0, dated 1/19/07IR

579658, Floating Head Removal from 3 'D' RHR RoomIR

579005,

RIS-9081 Causing HPSW Hi Rad AlarmAR A1599678,

RIS-9081 Causing HPSW Hi Rad AlarmAR A1599677,

RIS-9082 Causing HPSW Hi Rad AlarmC0220444 - '4T4' Bus; Inspect, Rework as RequiredA1605389 - '4T4' Bus Fault, Inspect, Rework as RequiredA1605391 - 3 'B' RFPT TrippedA1605414 - Loss of 30Y022-18A1605422 - 3 'A' Isophase Bus Cooling Fan Breaker TrippedA1605436 -

MO 3149B temporary powerA1605437 - 3 'B' D/W Chiller Trip

A-8AttachmentA1605471 - 3 'B' Isophase Bus FME InspectionAR A1607626 -

AO-2-23-042 Would Not Reopen During the Performance of

STST-O-023-301-2 - HPCI Pump, Valve, Flow and Unit Cooler Functional and In-service Test,Revision 47, completed 3/14/07Peach Bottom Operator Narrative Logs 3/14/07IR

604364, Human Error Results in E-3 EDG Overload & E-33 Breaker TripAR A1607776, Incomplete Procedure Performance During E-3 EDG Testing

Section 1R15: Operability EvaluationsIR 453260,

RHR to HPSW Leak - HPSW Sample Shows Radiological Contamination in 3 'B' LoopIR

583564, Unit 2 'B' Loop HPSW High Radiation AlarmIR

584041, RHR 3 'D' Heat Exchanger Lower Support Gap and Missing BoltIR

584070, Near Miss Opportunity for Potential 3.0.3 Inoperability AR A1551497-01, Assess Leak Rate Identified Via Bottom Head Sampling AR A1578288, Increased Leak Rate for 3 'D' RHR Heat ExchangerAR A1592631, 3 'D' RHR Exchanger/3 'B' RHR Loop Discharge Pipe FlushTRT 06-47, 3 'D' RHR Exchanger/3 'B' RHR Loop Discharge Pipe FlushECR

PB 96-03159-000, Emergency Cooling Tower Freezing Issue ECR

PB 96-03159-000, Attachment 1, Evaluation of Icing Conditions in the EmergencyCooling Tower ReservoirIR

593397, 2 'C' RHR Heat Exchanger Plug Insertion Tooling FailedAR A1546765-20, Evaluate Leaving Pop-A-Plug Tooling Inside Plugged Tube Peach Bottom Lost Parts DatabaseER-AA-2006, Lost Parts EvaluationsMA-AA-716-008, Attachment 9, Loss of Integrity Actions, Recovery from a Loss of FMEIntegrityMA-AA-716-008, Attachment 10, Loss of Integrity Notification and Recovery PlanIR

594481, RHR to HPSW Leakage Greater Than Acceptance CriteriaIR

148870, RHR Heat Exchanger Leak: Evaluate per CFRs and ODCMIR

372040, Suspected 2B RHR/HPSW Heat Exchanger HPSW In-LeakageAR A1607626 -

AO-2-23-042, Would not Reopen During the Performance of

STST-O-023-301-2, HPCI Pump, Valve, Flow and Unit Cooler Functional and In-service Test,Revision 47, completed 3/14/07Peach Bottom Operator Narrative Logs 3/14/07IR

00608000, Heat Transfer Test Unsat. Update PTRM EntryIR

513038, 3DE058 Requires Cleaning (Micro-fouling)IR

516995, 2DE058 Heat Transfer Test Unsat. Revise PTRM EntryA1577785, 3DE058 Requires Cleaning (Micro-fouling)RT-O-033-600-2, Revision 14, Flow Test of ESW to ECCS Coolers and Diesel GeneratorCoolersTRM 3.11 and Bases

A-9Attachment

Section 1R17: Permanent Plant ModificationsModificationsPB 02-00758, Add

SQUG Method for Seismic Qualification into UFSAR, Etc., Revision 0PB 03-00119, U3 Main and Unit Aux SPR Mod - Installation and Testing ECR, Revision

2PB-05-00068, E324 480V LV Bkrs - Replace OD Trip Devices with Solid State, Revision 0PB 05-00140, Replace Bearing Lube Oil Supply Ball Valves with Orifices, Revision 3PB 05-00155, Core Spray Line Break Detection Vulnerability, Revision 0PB-05-00159, Install Line Stop Hdwr to Replace

ESW 518 Valve, Revision 5PB 05-00195, P00507 U2 Power Range Neutron Monitoring Mod - Reactor Stability, Revision 0PB 05-00236, Revise HPSW Design Pressure in M-30, Issue calc

PM-1071, Revision 0CalculationsPM-1071, Calculation of Pressure Drop through HPSW System, Revision 0PM-1075, HPCI Lube Oil System Orifice Sizing, Revision 023-15SP, Pipe Stress Analysis and Support Evaluation for HPCI Lube Oil Line From Lube OilCooler 20E105, Revision 0Corrective Action Reports

21323

279193

294570

309624

485619

487311

558911

599882*600116*600132**NRC Identified During InspectionDrawingsE-911, Electrical Secondary and Control Conn MOV, Sheet 1, Revision 52E-359, Recirculation Pump Suction and Discharge Valve, Sheet 1, Revision 29E-1617, Single Line Meter and Relay diagram, Sheet 1, Revision 63MiscellaneousDPIS-2-14-043B Instrument Calibration Sheet, Revision 2Midas Calc Results, MOV

MO-2-02-053A, 10/2/06NE-164, Specification for Environmental Service Conditions Peach Bottom Atomic PowerStation Units 2 and 3, Revision 5P-T-17, Dynamic Qualification Program, Revision 4SQUG Letter, Revision 3A to the Generic Implementation Procedure for Seismic Verification ofNuclear Power Plant Equipment, dated 2/16/04SQUG

Memorandum, Use of GIP Revision 3A, dated 6/14/0533-55045-QS, Class 1E Electrical Equipment Environmental Qualification Report, Revision 26280-M1JJ-97, Instruction Manual Motor Operated Gate Valves, Revision 0

A-10Attachment11187-G-14, General Project Requirements for Seismic Design and Analysis of Equipment andEquipment Supports for Peach Bottom Atomic Power Station Units 2 & 3, Revision 0ProceduresAO 10.8-2, Placing Torus Cooling in Service with LOCA Signal Present or Has Occurred, Revision 8CC-AA-320-002, Use of SQUG Methodology for the Seismic Qualification of New andReplacement Items, Revision 0CC-AA-320-1004, Guidance for the Use of SQUG Methodology for the Seismic Qualification ofNew and Replacement Items, Revision 1M-055-005, 480 Volt I-T-E Solid State Breaker Trip Device Testing, Revision 1NE-C-420-04, Setpoint Methodology, Revision 1SE-10, Alternate Shutdown Procedure, Attachments 1-4, 7, Revision 14S0 48.1.B, Emergency Cooling Water System Startup, Revision 11Surveillance ProceduresST-O-054-753-2, E32 4KV Bus Undervoltage Relays and LOCA Loop Functional Test, Revisio

n 17Work Orders

A1188670 C0216690

Section 1R19: Post-Maintenance TestingA1602476,

ESW Pump 0AP057 Discharge Check ValveR1049544, ESW, Valve Unit Clr and ECT Fans

ISTST-O-033-300-2, Revision 31, ESW, Valve, Unit Cooler and ECT Fans Functional IST, performed 2/3/07ST-O-033-300-2, Revision 31, ESW, Valve, Unit Cooler and ECT Fans Functional IST,performed 2/4/07C0220132, 2-5A-K003A: Replace Relay and Perform PMTIR

00585972, 2-5A-K003A Relay FailedSI2M-60F-RT7-A4M2, Revision 4, Response Time Test of MSIV Closure Scram Channel AA1225120, Intake Struct Vent Exh 3AV83R0810095, E124-P-A (6244) Perform MCU InspectionAO 56.1, Revision 4, Removing and Installing a 480 VAC Motor Control Center BucketST-O-010-640-3, 3 'D' RHR Heat Exchanger Leak TestST-O-010-306-3, 'B' RHR Loop Pump, Valve, Flow and Unit Cooler Functional and InserviceTestA1607776, E-3

Diesel Generator, Incomplete Procedure Performance During Testing Resultsin E-3 Generator TripC0219643, 2AP040 Clean/Inspect/Repack Cylinders (2 'A' SLC Pump)ST-O-011-301-2, Standby Liquid Control Pump Functional Test for IST, completed 3/27/07

A-11AttachmentIR

0604364, E-3 Diesel Trip During TestingST-O-052-123-2, Diesel Generator RHR Pump Reject TestST-O-052-213-2, E3 Diesel Generator Slow Start Full Load and IST TestA1603535, U2 HPCI

ST-003 Modification PMT Unexpected ResultIR

00590626, U2 HPCI

ST-003 Modification PMT Unexpected ResultC0220288, Recal/Rework/Replace

LS-2-23-090 as Required (U2 HPCI Steam Supply Drain Pot Level)WO

C0220652, 0CG012-DR InspectionsWO R1011869,

CHK-O-33-515A; Disassemble Inspect/ReworkWO R0810095, E124-P-A (6244) Perform MCU Inspection

590973, Steam Leak through HV-2-23C-21173

Section 1R22: Surveillance TestingST-O-052-701-2, Rev 16,

E-1 Diesel Generator 24-Hour Endurance Test, completed 1/18/07SI3F-13-84-XXCQ, Revision 18, Calibration Check of RCIC Steam Line High Flow InstrumentDPIS, 3-13-84, completed 1/22/07SI3F-13-83-XXCQ, Revision 21, Calibration Check of RCIC Steam Line High Flow InstrumentDPIS, 3-13-83, completed 1/22/07ST-O-020-560-2, Reactor Coolant Leakage Test, Performed 1/27/07ST-O-033-300-2, Revision

31, ESW, Valve, Unit Cooler, and ECT Fans Functional IST,performed 2/4/07ST-O-010-301-3, 'A' RHR Loop Pump, Valve, Flow and Unit Cooler Functional and InserviceTest, performed 1/12/07ST-O-052-212-2, Revision 26, E-2 Diesel Generator Slow Start Full Load and IST Test,completed 1/24/07*IR

586970, UFSAR Table 4.8.1 Update on RHR Flow not Fully Encompassing*IR581062, DBD P-S-09 Not Updated for 3 'A' RHR Pump Motor ReplacementIR

559583, Apparent Conservative Error in Calc

ME-507IR

540115, Request for Engineering to Review Margin for 2 'D' RHR Pump Pressure/FlowDesign Basis Document (DBD) P-S-09, Residual Heat Removal SystemDesign Calculation Number

ME-0171, RHR Pump Discharge Pressure for Rated ConditionDesign Calculation Number

ME-0507, Acceptance Criteria for RHR Pumps Flow TestAmendment No. 27 to Facility Operating License No.

DPR-56, Docket 50-278, dated November 15, 1976ECR No.

PB-99-00079-000, Discrepancy Identified During Review of UFSAR Section 4.4 & 6.3Engineering Work Request (EWR) P-51688, ST Requirements for RHR PumpsEWR P-51497, Unit 3 RHR System CalculationsEWR P-50900, ST Requirements for RHR PumpsSI3F-23-82-XXC2, Calibration Check of HPCI Flow Instruments FT 3-23-082, FI/FC 3-23-108,E/S 3-23-143, XS 3-23-144 and FS 3-23-078, Revision 3, performed 3/20/07Technical Specifications 3.3.5.1.4, 3.3.5.1.5 and 3.5.1*Identified as a result of this inspection

2Attachment

Section 1R23: Temporary Plant ModificationsECR

PB 07-00080, Temporary Power for 30Y023Drawing E-1700, Revision 38, sheet 1IR

00596812, Both LEFM Computers De-energized Due to Loss of 30Y023IR

00596818, Temporary Power for 30Y023ECR

PB 07-00081, Temp Power for 4-T-4-T-CDrawing E-1700, Revision 38, sheet 1WO

C0220453, Provide Temp Power to

MO-3-06C-3149BWO

C0220454, Provide Temp Power to 30Y022

Section 1EP6: Drill EvaluationIR 580462,

DEP PAR Failure Section 2PS1: Radioactive Gaseous and Liquid Effluent Treatment and MonitoringSystemsDocuments2005 Radioactive Effluent Release Report No. 48, dated April 25, 2006, (including Projected Public Dose Assessments);2004 Radioactive Effluent Release Report No. 47, dated April 27, 2005, (including Projected Public Dose Assessments;2005 Radiation Dose Assessment Report No. 21, dated April 25,

20062004 Radiation Dose Assessment Report No. 20, dated April 29, 2005Changes to Offsite Dose Calculation Manual and Technical Justifications for ODCM ChangesSelected 2004, 2005, 2006 Analytical Results for Radioactive Liquid, Charcoal Cartridge,Particulate Filter, and Noble Gas Samples Implementation Records for the Compensatory Sampling and Analysis Program when theEffluent Radiation Monitoring System (RMS) is Out-of-Service Calibration Records for Chemistry Laboratory Measurements Equipment (Gamma)Implementation Records of the Measurement Laboratory Quality Control Program, IncludingControl ChartsImplementation Records of the Intra-laboratory Comparisons by the Licensee and theContractor Laboratory

==Section 4OA3: Event FollowupIR

554800, Potential External Flood Vulnerability Found for==

EDG BuildingIR

558326, Diesel Building's Oil Separator Pit Check Valve Needs InspectionIR

570723, Circulating Water Pump Structure Flood Program VulnerabilityIR

522005, Inspect EDG Room Equipment Drain Backwater ValvesIR

523285, Improvements to Plant Response to External Flood (RE: EDGS)IR

505423, Emergency Diesel Building Flooding - Check Valve and IPE IssuesIR

534622, Multiple MSIV LLRT Failures: P2R16

A-13AttachmentIR

539591, Review/Approval of FMCT for 80D Inboard MSIV not DocumentedIR

539594, New Main Poppet Used for

MSIV-80D Dimensionally DifferentIR

539633,

AO-2-01A-080D Had Unsat Blue Check After Poppet ReplacementIR

539186, Temporary Change to MSIV LLRT Procedure InadequateIR

538998,

AO-2-01A-08D Failed AS-left LLRT, Rework RequiredIR

534610, Discrepancies in U2 MSIV (86A, 86B & 86D) LLRT ResultsIR

539527, NOS ID - MSIV Hit Not IAW Troubleshooting ProcedureIR

540128, Seat Polishing of MSIVs - Improvement OpportunityIR

563253, External Flood Vulnerability - Circulating Water Pump StructureIR

554800, External Flood Vulnerability Found for EDG BuildingIR

520322, E-3 EDG Fire at Roof Exhaust PenetrationIR

604364, Incomplete Procedure Performance During E-3 Diesel TestingST/LLRT 20.01A.02, Revision 6, Main Steam Isolation Valve Local Leak Rate TestSpecial Event Procedure (SE)-4, Flood, Revision 21

ST-O-052-123-2, E-3 Diesel Generator RHR Pump Reject Test, Revision 4ST-O-054-951-2, Offsite and Onsite Electrical Power Breaker Alignment and Power AvailabilityCheck with a Start-up Source and/or EDG Inoperable, Revision 6SO 52A.1.B, Diesel Generator Operations, Revision 38Quick Human Performance Investigation, Missed Procedure Step Results in Unplanned E-3EDG Load Change and E-33 Breaker TripAR

1607776, Incomplete Procedure Performance During E-3 EDG TestingPBAPS Operations Standing Order, 07-01, Peer Check Standards Clarifications and Expectations, 3/22/2007

IR 596616, Fault AT PB 3 50D E CBM '4T4' (0264) 3 'B' Iso-Phase Cooler FanIR

596767, Fire Brigade Critique Following U3 Breaker FireIR

597185, Drywell Chilled Water Not Modeled in PRA, Nor in ParagonIR

597214, LTA Guidance to Determine High Risk Evolution (HRE) in ParagonIR

597308, Security Critique Enhancement from 02/27/07 UE EventIR

597381, Nos ID: Opportunity for Improved '4T4' QuarantineIR

597402, Evaluate Recirc Pump MismatchIR

596889, UE Declared for Unit 3 Due to a Fire in the '4T4' LCIR

598869, Hole on the Side of Breaker Cubical (FME)IR

599184, Extend of Condition Walkdown of U2 480L LC BusIR

601094, Failure to Contact OEM to Repair '4T4' 480V Load CenterIR

601326, 30Y022 Panel Circuit 20 Won't Stay EnergizedIR

606397, Perform ITE Rejection Tab WalkdownIR

521321, ENS Communicator Issues During 8/15/06 EDG UEFire Event Report, Peach Bottom/Unit 3, 02/27/2007Event Number: 43189, UE Fire Inside the Unit 3 Turbine Area Load Center, 02/27/2007 Preliminary Notification of Event or Unusual Occurrence -

PNO-I-07-002, Notification ofUnusual Event (NOUE) Declared Due to Fire in Turbine Building Load Center at Peach BottomUnit 3, February 27, 2007

A-14Attachment

LIST OF ACRONYMS

ADAMSAgency-wide Documents Access and Management SystemALARAas low as reasonably achievableARaction requestAVapparent violationCAPcorrective action programCDFcore damage frequencyCFRCode of Federal RegulationsCVcontainment ventingDBDDesign Basis DocumentDEPdrill & exercise performanceDRPDivision of Reactor ProjectsEALemergency action levelECTemergency cooling towerEDGemergency diesel generatorEOPsemergency operating proceduresESWemergency service waterFBfire brigadeHXheat exchangerHPCIhigh pressure coolant injectionHPSWhigh pressure service waterIMCInspection Manual ChapterINInformation NoticeIPInspection ProcedureIPEIndividual Plant ExaminationIRissue reportISTinservice testJPMsjob performance measureskVkilovoltLERslicensee event reportsLERFlarge early release frequencyLOCAloss-of-coolant accidentMCRmain control roomMRMaintenance RuleMSIVsmain steam isolation valvesNCVnoncited violationNEINuclear Energy InstituteNRCNuclear Regulatory CommissionNRRNuclear Reactor RegulationODCMOffsite Dose Calculation ManualPARprotective action recommendationPARSPublicly Available RecordsPBAPSPeach Bottom Atomic Power StationPIperformance indicator

A-15AttachmentPMTpost-maintenance testingQAquality assuranceRCICreactor core isolation coolantRCSreactor coolant systemRCWPreactor cavity work platformREMPradiological environmental monitoring programRETSRadiological Effluent Technical SpecificationsRGRegulatory GuideRHRresidual heat removalROsreactor operatorsRTPrated thermal powerSDPsignificance determination processSEssafety evaluationsSOsenior operatorSPCsuppression pool coolingSRAsenior reactor analystSROsenior reactor operatorSRVsafety relief valveSSCstructure, system, and component

ST urveillance testSWRssimulator work requestsTRMTechnical Requirements ManualTSTechnical SpecificationUEunusual eventURIunresolved itemUFSARUpdated Final Safety Analysis ReportWECSwork execution control supervisorWOwork order