IR 05000277/2007003

ESW Out-of-Service. b.FindingsNo findings of significance were identified..2Complete System Walkdown (71111.04S - 1 Sample) a.Inspection ScopeDuring the week of June 25, 2007, the inspectors performed one complete Unit 2 highpressure service water (

PB-108-111-1001, "Preparation for Severe Weather," increased

the online risk assessment to "Yellow," and subsequently implemented procedure

AO 53.2-0, "Equipment Checks After a Thunderstorm." b.FindingsNo findings of significance were identified.1R04Equipment Alignment (71111.04Q - 3 Partial Walkdown Samples).1Partial Walkdown a.Inspection ScopeThe inspectors performed a partial walkdown of three systems to verify the operability ofredundant or diverse trains and components when safety-related equipment was

inoperable. The inspectors performed walkdowns to identify any discrepancies that could

impact the function of the system and potentially increase risk. The inspectors reviewed

applicable operating procedures, walked down system components, and verified that

selected breakers, valves, and support equipment were in the correct position to support

system operation. The inspectors also verified that

PBA [[]]

PS had properly identified and

resolved equipment alignment problems that could cause initiating events or impact the

capability of mitigating systems or barriers and entered them into the CAP. The three

systems reviewed were: *E-3 Diesel Generator and 3 Startup Transformer with the 2 Startup TransformerOut-of-Service;*Unit 2 Reactor Core Isolation Cooling (RCIC) with Unit 2 High Pressure CoolantInjection (HPCI) Out-of-Service; and*'B' Emergency Service Water (ESW) with 'A'

HPSW) system walkdown of the accessible portions of the

3Enclosuresystem. The full walkdown was performed to identify any discrepancies which couldimpact the Unit

2 HP [[]]

SW system function. The inspectors reviewed system operating

procedures, piping and instrumentation drawings, walked down control system

components, and verified that circuit breakers and valves were in the appropriate

positions. b.FindingsNo findings of significance were identified.1R05Fire Protection (71111.05Q - 10 Samples)Fire Protection - Tours a.Inspection ScopeThe inspectors reviewed

PBAPS 's Fire Protection Plan, Technical Requirements Manual(

TRM), and the respective pre-fire action plan procedures to determine the required fire

protection design features, fire area boundaries, and combustible loading requirements for

the areas examined during this inspection. The fire risk analysis was reviewed to gain risk

insights regarding the areas selected for inspection. The inspectors performed

walkdowns of ten areas to assess the material condition of active and passive fire

protection systems and features. The inspection was also performed to verify the

adequacy of the control of transient combustible material and ignition sources, the

condition of manual firefighting equipment, fire barriers, and the status of any related

compensatory measures. The following ten fire areas were reviewed for impaired fire

protection features:*Unit 3 Reactor Building (RB),

RC [[]]

IC Room, 88' Elevation (Fire Zone 63);*Standby Gas Treatment Room, Radwaste Building, 91'6" Elevation (Fire Zone 70);

Unit

3 RB , North Control Rod Drive (

CRD) Equipment and West Corridor (Fire Zone 13H);*Unit 3 Refuel Floor (Fire Zone 55);

Unit 3 'A' and 'C' Core Spray Rooms (Fire Zones 13D & 13E);

Unit 2 Emergency Battery/Switchgear Rooms (Fire Zone 127);

Unit

2 RC [[]]

IC (Fire Zone 60);

Unit 2 Main Transformer Yard (Fire Zone 151);

2 Startup Switchgear Building (Fire Zone 164); and

Diesel Generator Building, 127' Elevation (Fire Zone 132). b.FindingsNo findings of significance were identified.

4Enclosure1R11Licensed Operator Requalification Program (71111.11Q - 1 Sample) Resident Inspector Quarterly Review a.Inspection ScopeOn June 12, 2006, the inspectors observed operators in

PBA [[]]

PS's simulator duringlicensed operator requalification training to verify that operator performance was adequate

and that evaluators were identifying and documenting crew performance issues. The

inspectors verified that performance issues were discussed in the crew's post-scenario

critiques. The inspectors also observed operator implementation of procedures. The

inspectors discussed the training, simulator scenarios, and critiques with the operators,

shift supervision, and the training instructors. The evaluated scenario observed for this

one sample involved the events listed below: *Small Break Loss of Coolant Accident; and*An Anticipated Transient Without Scram. b.FindingsNo findings of significance were identified.1R12Maintenance Effectiveness (71111.12Q - 2 Samples) a.Inspection ScopeThe inspectors reviewed two samples of

PBAPS 's evaluation of degraded conditionsinvolving safety-related structures, systems, and components (

SSCs) for maintenance

effectiveness during this inspection period. The inspectors reviewed

PBA [[]]

PS's

implementation of the Maintenance Rule (MR), and verified that the conditions associated

with the referenced condition reports (CRs) were evaluated against applicable MR

functional failure criteria as found in the licensee's scoping documents and procedures.

The inspectors also discussed these issues with system engineers and MR coordinators

to verify that they were tracked against performance criteria and that the systems were

classified in accordance with MR implementation guidance. Documents reviewed during

the inspection are listed in the Attachment. The following conditions were reviewed:*Issue Report (IR) 587171,

ESW Check Valve (

CHK -0-33-515A) - Not SeatedCauses

ESW [[]]

ST-O-033-300-2 to be Aborted; and*IR 622560, Maintenance Preventable Functional Failure for Loss of '4T4' 480 VoltLoad Center. b.FindingsNo findings of significance were identified.

5Enclosure1R13Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 Samples) a.Inspection ScopeThe inspectors evaluated

PBAPS 's implementation of their maintenance risk program toverify that

PBAPS managed risk in accordance with 10 CFR Part 50.65(a)(4). Procedure

WC -

AA-101, "On-line Work Control Process," was also reviewed. This inspection

included reviews of

PBA [[]]

PS's use of the Paragon online risk monitoring software. The

inspectors reviewed equipment tracking documentation, daily work schedules, and

performed plant tours. The following activities selected were based on plant maintenance

schedules and systems that contributed to risk. The inspectors completed eight

evaluations of maintenance activities on the following:*Troubleshooting, Rework and Testing (TRT) Control Form No. 07-18, Monitor3 'A' Recirculation Pump Seal Parameters During Recirculation Pump Speed

Changes;*TRT No.07-020, Re-align

CRD Pump Suction to the Condensate Storage Tank(

CST) from the Condensate System;*WO C0220911, Calibrate, Repair & Replace E-2 EDG Temperature Switch;

WO A1613202, 3 'B' Recirculation Pump 2nd Stage Seal Pressure;*IR 623723, Bolt and Heli-coil Found Damaged at Disassembly on 00T634;

IR 626534, Equipment Not Protected As Required;

WO R0736769-01, Core Spray Loop 'A' Full Flow Test Valve Operator,MO-2-14-026A-OP, Perform Motor Operator Preventive Maintenance; and*IR 542109, 2 'C' Service Air Compressor Trip.Additionally, the inspectors verified that an inspector-identified issue,

IR 626534,"Equipment Not Protected As Required," was entered into the

PBAPS's CAP. b.FindingsNo findings of significance were identified.1R15Operability Evaluations (71111.15 - 5 Samples) a.Inspection ScopeThe inspectors reviewed five issues to assess the technical adequacy of the evaluations,the use and control of compensatory measures, and compliance with the licensing and

design bases. Associated adverse condition monitoring plans, engineering technical

evaluations, and operational and technical decision making documents were also

reviewed. The inspectors verified these processes were performed in accordance with

the applicable procedures. The inspectors used

TS ,

TRM, the Updated Final Safety

Analysis Report (UFSAR), and associated Design Basis Documents (DBDs) as referencesduring these reviews. The issues reviewed included:*Non-Safety Related Piece Installed in E-4

EDG Part (

IR 615413);*Rising 3 'A' Reactor Recirculation Pump (RRP) #2 Seal Temperature (IR 617988);

Provide Supplemental Cooling to the 3 'A'

RRP Seal (

IR 618478);

6EnclosureTarget Rock Safety/Relief Valve (SRV) Seal Welds: Potential Code Issue(IR 628251); andSmall Leak on 2 'B' Main Steam Line Differential Pressure Instrument LineSnubber Threaded Cap (IR 627026). b.FindingsNo findings of significance were identified.1R19Post-Maintenance Testing (71111.19 - 7 Samples) a.Inspection ScopeThe inspectors observed selected portions of post-maintenance testing (PMT) activitiesand reviewed completed test records. The inspectors observed whether the tests were

performed in accordance with the approved procedures and assessed the adequacy of

the test methodology based on the scope of maintenance work performed. In addition,

the inspectors assessed the test acceptance criteria to evaluate whether the test

demonstrated that the tested components satisfied the applicable design and licensing

bases and the TS requirements. The inspectors reviewed the recorded test data to verify

that the acceptance criteria were satisfied. The inspectors reviewed seven PMTs

performed in conjunction with the following maintenance activities:*WO C0220911, Calibrate, Repair & Replace E-2

EDG Temperature Switch;*

WO R1049367, Unit 3 Hydraulic Control Unit (HCU) 50-43: HCU Overhaul;

WO R1017055,

DDFP (00P063-

DR) Diesel Engine 6YR Overhaul;

WO C0216504,

RCIC Suction Pressure Switch (

PS-2-13-067-01), Replace Pressure Switch;

WO C0221445, Inspect/Repair/Replace Unit 2 'C' Main Steam Line Radiation

Monitor (RIS-2-17-251C);*WO C0215740, Replace Unit 2 'B' Reactor Protection System Motor Generator Set Endbell; and*WO R0629147, Perform Motor Control Unit Inspection on the 'C' Glycol Pump. b.FindingsNo findings of significance were identified.1R22Surveillance Testing (71111.22 - 5 Samples) [3 Routine Samples;

1 IST Sample; 1Reactor Coolant System (

RCS) Leakage Sample] a.Inspection ScopeThe inspectors reviewed and observed portions of selected surveillance tests (STs), andcompared test data with established acceptance criteria to verify the systems

demonstrated the capability of performing the intended safety functions. The inspectors

also verified that the systems and components maintained operational readiness, met

applicable TS requirements, and were capable of performing the design basis functions.

The five STs reviewed and observed included:

7Enclosure*ST-O-023-301-3,

HPCI [[Pump, Valve, Flow and Unit Cooler Functional andIn-Service Test []]

IST Sample];*ST-O-020-560-2 & 3, Reactor Coolant Leakage Test [RCS Leakage Sample];

ST-I-60A-230-3, Linear Power Range Monitor Gain Calibration;

SI2T-MIS-8547-C1CQ, Calibration/Functional Check of Channel 'C' Group 1, 4and 5 of Primary Containment Isolation Valve (PCIV) Logic for

TS s-80547C; and*

ST-R-003-485-3, CRD Scram Insertion Timing of Selected Control Rods.b.FindingsNo findings of significance were identified.1R23Temporary Plant Modifications (71111.23 -1 Sample) a.Inspection ScopeThe inspectors reviewed one temporary modification to verify that implementation of themodification did not place the plant in an unsafe condition. The review was also

conducted to verify that the design bases, licensing bases, and performance capability ofrisk significant SSCs had not been degraded as a result of the modification. The

inspectors verified the modified equipment alignment through control room

instrumentation observations; the

UFSAR ; drawings; procedures;

WO reviews; and plant

walkdowns of accessible equipment. The following temporary modification was reviewed:*TCCP 07-00172, Install Cooling Unit to Assist 3 'A'

RRP Seal Cooling. b.FindingsNo findings of significance were identified.Cornerstone: Emergency Preparedness1

EP 2Alert and Notification System (ANS) Evaluation (71114.02 - 1 Sample) a.Inspection Scope An onsite review was conducted to assess the maintenance and testing of the

ANS testing and maintenance.

ANS. During this inspection, the inspectors interviewed emergency preparedness (EP)

staff responsible for implementation of the

IRs pertaining

to the ANS were reviewed for causes, trends, and corrective actions. The inspectors

further discussed with

PBAPS , the

ANS siren system and its performance from July 2005

through May 2007. The inspectors reviewed the licensee's procedures and the ANS

design report to ensure compliance with those commitments for system maintenance and

testing. The inspection was conducted in accordance with

NRC Inspection Procedure (

IP)

71114, Attachment 2. Planning standard, 10 CFR 50.47(b)(5) and the related

requirements of 10 CFR 50, Appendix E were used as reference criteria.

8Enclosure b.FindingsNo findings of significance were identified.1EP3Emergency Response Organization (ERO) Staffing and Augmentation System (71114.03 - 1 Sample)

a.Inspection Scope A review of Peach Bottom's

ERO augmentation staffing requirements and the process fornotifying the

ERO was conducted. This was performed to ensure the readiness of key

staff for responding to an event and to ensure timely facility activation. The inspectors

reviewed procedures and

IR s associated with the

ERO notification system and drills, and

reviewed records from call-in drills. The inspectors interviewed personnel responsible for

testing the ERO augmentation process, and reviewed the training records for a sampling

of the ERO to ensure training and qualifications were up-to-date. The inspectors reviewed

procedures for

ERO administration and training, and verified a sampling of the

ERO

participated in exercises in 2005 and 2006. The inspectors also reviewed records of

offsite agency training and the June 2007 Respirator Qualification Report. The inspection

was conducted in accordance with

NRC [[]]

IP 71114, Attachment 3. Planning standard,

CFR 50.47(b)(2) and related requirements of 10

CFR 50, Appendix E were used as

reference criteria. b.FindingsNo findings of significance were identified.

1EP 4Emergency Action Level (

EAL ) and Emergency Plan Changes (71114.04 - 1 Sample) a.Inspection Scope Since the last

NRC inspection of this program area, Emergency Plan (Plan), Revision 26,was implemented based on

PBAPS's determination, in accordance with

CFR 50.54(q), that the changes resulted in no decrease in effectiveness of the Plan,

and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and

Appendix E to 10 CFR 50. The inspectors conducted a sampling review of the

Emergency Plan changes, and changes to the lower-tier Emergency Plan implementing

procedures, to evaluate the changes for potential decreases in effectiveness of the

Emergency Plan. However, this review was not documented in a safety evaluation report

and does not constitute formal NRC approval of the changes. Therefore, these changes

remain subject to future NRC inspection in their entirety. b.FindingsNo findings of significance were identified.

9Enclosure1EP5Correction of Emergency Preparedness Weaknesses (71114.05 - 1 Sample) a.Inspection Scope The inspectors reviewed a sampling of self-assessment procedures and reports to assessPBAPS's ability to evaluate their performance and programs. The inspectors reviewed a

sampling of IRs from July 2006 through May 2007, initiated by Exelon Nuclear at Peach

Bottom from drills, self-assessments, and audits. Other drill reports reviewed included:

medical/health physics, fire, integrated, and call-in. Additionally, the inspectors reviewed

the three UE Evaluation Reports generated since the last inspection, and audits for 2006

and 2007 required by 50.54(t). This inspection was conducted in accordance with

NRC [[]]

IP

71114, Attachment 5. Planning standard, 10 CFR 50.47(b)(14) and the related

requirements of

10 CFR 50, Appendix E were used as reference criteria. b.FindingsNo findings of significance were identified.1

EP6Drill Evaluation (71114.06 - 1 Sample)Off-Year Exercise (Drill) a.Inspection ScopeThe inspectors conducted this inspection to assess: training quality and conduct;emergency plan procedure implementation; facility and equipment readiness; personnel

performance in drills and exercises; organizational and management changes; and

communications equipment readiness. The primary focus of this inspection was to verify

PBA [[]]

PS's critique of classification, notification, and protective action recommendation

(PAR) development.On May 15, 2007, the inspectors observed a full scale drill. The primary focus of thisinspection was to verify

PBAPS 's critique of classification, notification, and

PAR

development. Selected portions of the drill were observed in the control room simulator

and later in the technical support center (TSC). The drill scenario began with a simulated

internal flooding event in the 2 'A' residual heat removal (RHR) pump room that degraded

the performance of the associated safety system. The drill scenario continued with a

simulated reactor event that started with a reduction of coolant flow to the core and

progressed until three fission product barriers (fuel cladding, RCS, and containment) were

lost. The inspectors observed licensed operator and ERO personnel adherence to the

Emergency Plan implementing procedures. The ERO personnel responses to simulated

degraded plant conditions were inspected to identify weaknesses and deficiencies in

classification and notification. The inspectors also observed the transition of responsibility

for the

ERO from the shift manager in the simulated control room to the

TSC. The

inspectors observed

PBAPS 's critique of the drill to evaluate

PBAPS's identification of

weaknesses and deficiencies. The inspectors compared

PBA [[]]

PS's identified issues

against the inspectors' observations to determine whether

PBA [[]]

PS adequately identified

problems and entered them into the CAP. This inspection activity represented one

10Enclosuresample. The documents and procedures reviewed during the inspection are listed in theAttachment. b.FindingsNo findings of significance were identified.2.RADIATION

SAFETY Cornerstone: Public Radiation Safety 2

PS 2Radioactive Material Processing and Transportation (71122.02 - 5 Samples).1Inspection Planning/In-Office Inspection a.Inspection ScopeThe inspectors reviewed the solid waste system description in the

UFS [[]]

AR and recentradiological effluent release reports for information on the types and amounts of

radioactive waste. The inspectors reviewed Exelon's audit program in the area of radioactive wastecharacterization, transportation, and disposal. The inspectors also reviewed the status of

the

NRC approved quality assurance program in this area. (Section 2

PS2.6) b.FindingsNo findings of significance were identified. .2Radioactive Waste System Walkdown a.Inspection ScopeThe inspectors walked down accessible portions of the station's radioactive liquid andsolid waste collection, processing, and storage systems and locations to determine if:

systems and facilities were consistent with descriptions provided in the

UFS [[]]

AR; to

evaluate their general material conditions; and to identify changes made to systems.

Areas visually inspected included tank and pump rooms, the de-watering facility, in-plant

and outside waste storage areas, outside tank areas, and the low level-waste storage

facility. Visual inspection records and previous surveys were also reviewed. The

inspector also discussed operation of the radwaste systems with cognizant licensee

personnel.The inspectors reviewed the status of any non-operational or abandoned radioactivewaste process equipment; the adequacy of administrative and physical controls for those

systems; changes made to radioactive waste processing systems and potential

radiological impact, including conduct of safety evaluations of the changes, as necessary.

11EnclosureThe inspectors reviewed the current processes for transferring radioactive waste resin andsludge to shipping containers and mixing and sampling of the waste, as appropriate, to

evaluate waste mixing, adequacy of sampling, and the methodology for waste

concentration averaging. The inspector also reviewed radioactive waste and material

storage and handling practices; sources of radioactive waste at the station (waste

streams) and processing (as appropriate) and handling of the waste; and the general

condition of facilities and equipment. The review was against criteria contained in the station's

UFSAR , 10

CFR Part 20,10

CFR 61, the Process Control Program (

PCP), and applicable station procedures. b.FindingsNo findings of significance were identified..3Waste Characterization and Classification a.Inspection ScopeThe inspector reviewed the following matters:

Radio-chemical sample analysis results for radioactive waste streams;*Development of scaling factors for difficult to detect and measure radionuclides;

Methods and practices to detect changes in waste streams;

Classification and characterization of waste relative to

10 CFR 61.55 and 10

CFR 61.56;*Implementation of applicable

NRC branch technical positions (

BTPs) on wasteclassification, concentration averaging, waste stream determination, and sampling

frequency;*Current waste streams and their processing relative to descriptions contained inthe

UFSAR and the station's approved

PCP ; *Current processes for transferring radioactive waste resin and sludge dischargesinto shipping/disposal containers to determine adequacy of sampling; *Revisions of the

PCP and the

UFSAR to reflect changes (as appropriate); and

Waste processing topical report (de-watering).The inspector discussed the adequacy of samples collected from the waste transfer andde-watering system.The review was against criteria contained in

NRC BTPs, and Exelon procedures. b.FindingsNo findings of significance were identified.

2Enclosure.4Shipment Preparation a.Inspection ScopeThe inspector observed a non-exempt radioactive material shipment (PM-07-057) inpreparation. The inspector reviewed associated transportation documents, reviewed

radiological surveys to support transportation, reviewed license requirements, and

discussed preparation with cognizant Exelon personnel. The inspector also reviewed

personnel training relative to

NRC Bulletin 79-19 and 49

CFR 172, Subpart H. The

inspector reviewed and discussed technical training presented to workers. The inspector

verified that a training program was provided to personnel responsible for the conduct of

radioactive waste processing and radioactive waste shipping activities. b.FindingsNo findings of significance were identified. .5Shipment Records and Documentation a.Inspection ScopeThe inspector selected and reviewed the records associated with six non-exceptedshipments of radioactive material made since the previous inspection in this area

(Shipment Nos.

PM -07-057,

PW -07-010,

PW -06-030,

PW-07-007, PW-07-001,

PW-07-003). The shipments were selected based on waste classification and

waste-stream characteristics. The following aspects of the radioactive waste, radioactive

material packaging, and radioactive material shipping activities were reviewed:*Implementation of applicable shipping requirements including completion of waste manifests;*Implementation of the specifications in applicable Certificates of Compliance, asappropriate, for the approved shipping casks including limits on package contents;*Classification and characterization of waste relative to

10 CFR 61.55 and 61.56, asappropriate;*Implementation of up-to-date

NRC and Department of Transportation (DOT)shipping requirements;*Implementation of 10 CFR 20, Appendix G;

Implementation of specific radioactive material shipping requirements;

Packaging of shipments;

Labeling of shipping containers;

Placarding of transport vehicles;

Conduct of vehicle checks;

Provision of driver exclusive use and emergency instructions, as applicable;

Completion of shipping paper/disposal manifest;

Evaluation of package against package performance standards, as appropriate;

Conformance with procedures for cask loading, closure and use requirementsincluding consistency with cask vendor approved procedures; and*Use of latest revision documents.

13EnclosureThe review was against criteria contained in

10 CFR 20; 10

CFR 61;

10 CFR 71;applicable

DOT requirements, as contained in 49 CFR 170-189 for the above areas;

station procedures; applicable disposal facility licenses; and applicable Certificates ofCompliance or vendor procedures for various shipping casks.The inspector also selectively reviewed the 2006 Annual Radioactive Effluent ReleaseReport, relative to types and quantities of radioactive waste shipped offsite and relative to

changes to the

PCP. [[b.FindingsNo findings of significance were identified..6 Audits and Assessments of Radioactive Waste Handling a.Inspection ScopeThe inspector reviewed audits and assessments of the radioactive waste handling,processing, storage, and shipping programs, including the]]

PCP. The inspector also

reviewed selected corrective action documents written since the previous inspection. The

following documents were reviewed:*Chemistry, Radwaste, and Process Control Audit, (NOSA-PEA-06-04 (IR 476157),May 3, 2006; *Self-Assessment,

AS [[]]

SA-565928 A05, May 14, 2007; and

Issue Reports (IRs) 632879, 626897, 626873, 618653, 612012, 605803,592478486694, 240959, 642483, 642097, 642491, 632526, 486694. The review was against criteria contained in

10 CFR 20 Appendix G, 10

CFR 71.101, andapplicable station audit and surveillance procedures. b.FindingsNo findings of significance were identified.4.OTHER

ACTIVI [[]]

TIESC ornerstones: Barrier Integrity & Emergency Preparedness

4OA 1Performance Indicator (

PI) Verification (71151 - 7 Samples)

.1Barrier Integrity

PI s ( 71151 - 4 Samples) a.Inspection ScopeThe inspectors reviewed a sample of

PBAPS 's submittals for the four Barrier Integrity

PI slisted below to verify the accuracy of the data reported. The

PI definitions and the

guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment

Indicator Guideline," Revision 4, and Exelon procedure

LS -

AA-2001, "Collecting and

14EnclosureReporting of

NRC Performance Indicator Data," were used to verify that the reportingrequirements were met. The inspectors reviewed raw

PI data collected since

January 2006 to December 2006 and compared graphical representations from the most

recent

PI report to the raw data to verify the data was included in the report. The

PIs

reviewed were:*Unit 2 and Unit

3 RCS Specific Activity; and*Unit 2 and Unit 3

RCS Leakage. b.FindingsNo findings of significance were identified..2Emergency Preparedness (EP)

PI s (71151 - 3 Samples) a.Inspection ScopeThe inspectors reviewed data for the following

EP PIs:

Drill and Exercise Performance (DEP);*ERO Drill Participation; and

ANS Reliability. The inspectors reviewed supporting documentation from drills and tests from April 2006through March 2007, to verify the accuracy of the reported data. The review of these PIs

was conducted in accordance with

NRC [[]]

IP 71151. The acceptance criteria used for the

review were

10 CFR 50.9 and

NEI 99-02, Revision 4, "Regulatory Assessment

Performance Indicator Guidelines." b.FindingsNo findings of significance were identified.4OA2Identification and Resolution of Problems (71152 - 1 Sample) .1Routine Review of Items Entered Into the

CAP a.Inspection ScopeAs required by

IP 71152, "Identification and Resolution of Problems," and in order to helpidentify repetitive equipment failures, human performance issues or program issues for

follow-up, the inspectors performed routine screening of issues entered into

PBA [[]]

PS 's

CAP. This review was accomplished by selectively reviewing copies of

IR s and accessing

PBA [[]]

PS's computerized database. b.FindingsNo findings of significance were identified.

15Enclosure.2Review of Operator Work-Arounds (OWAs) (71152 - 1 Work-Around Sample) a.Inspection ScopeAs required by

IP 71152, "Identification and Resolution of Problems," the inspectorsconducted a review of the

OWA program to verify that

PBAPS was identifying

OWAs

problems at an appropriate threshold, have entered them in the CAP, and proposed or

implemented appropriate corrective actions. The inspectors reviewed the list of OWAs

and operator challenges (OCs) identified and managed in accordance with Exelon

procedure,

PBAPS *Unit 3 Steam Jet-Air Ejector (

AA-102-103, "Operator Work-Around Program." Specifically, the review

was conducted to determine if any OWAs for mitigating systems affected the mitigating

system's safety functions or affected the operator's ability to implement abnormal and

emergency operating procedures. The inspectors reviewed the following open OWAs

being tracked by

SJAE) Suction Valves Fail to Open When Placing theSJAE In-Service (Action Request (AR) A1536806).The inspectors also reviewed the lists of open OCs (deficiencies that are obstacles tonormal plant operations), periodically walked down the panels in the main control room,

and reviewed control room deficiencies to identify and be cognizant of: (1) OWAs that

have not been evaluated by

PBAPS , and (2)

OWAs that increase the potential for

personnel error, including OWAs that: *Require operations contrary to past training or require more detailed knowledge ofthe system than routinely provided; *Require a change from longstanding operational practices;

Require operation of a system or component in a manner dissimilar from similarsystems or components;*Create the potential for the compensatory action to be performed on equipment orunder conditions for which it is not appropriate;*Impair access to required indications, increase dependence on oralcommunications, or require actions under adverse environmental conditions; and*Require the use of equipment and interfaces that had not been designed withconsideration of the task being performed.

b.FindingsNo findings of significance were identified..3Semi-Annual Review to Identify Trends (71152 - 1 Semi-annual Trend Sample) .aInspection ScopeAs required by

IP 71152, Identification and Resolution of Problems, the inspectorsreviewed a list of approximately 5,000

IR s that Exelon initiated at

PBA [[]]

PS from December

1, 2006 through June 1, 2007, to perform the semi-annual PI&R trend review.

Approximately, 30 IRs were reviewed in detail to verify that the issues were adequately

identified, appropriately evaluated and corrected. The inspectors review was focused on

16Enclosurehuman performance issues. The review also included issues documented within

PBA [[]]

PS'sStation Trend Review for the fourth quarter of 2006 and the first quarter of 2007. b.Assessments and ObservationsAlthough no findings of significance were identified, the inspectors observed that the plantis being challenged by human performance deficiencies. Specifically, procedure

adherence was the aspect of human performance that was most frequently challenged.

Examples are documented in IRs 568038, 577381, 581258, 604364, 596616, 626534 and

633532. Procedure quality was another aspect of human performance that was

challenged. Examples are documented in IRs 635028, 633532, and 600686. However,

the inspectors did not identify any new trends that were not previously identified by

PBA [[]]

PS under their quarterly Station Trend Review reports. The inspectors noted that the

Station Trend Review report had identified procedure adherence issues as an emerging

trend. The inspectors also noted that improving human performance was identified as

one of five Station Focus areas for 2007.4OA3Event Followup (71153 - 5 Samples) .1(Closed) Unresolved Item (URI)05000277/2007002-04, Incorrect Size Breaker Resultedin a Fire in the '4T4' 480 Volt Load Center a.Inspection ScopeURI 05000277/2007002-04 was opened in

NRC Inspection Report 050000277;05000278/2007002.

PBAPS had preliminarily determined that the fire resulted from an

apparent mismatch between the ratings of one breaker and its cubicle in the '4T4' 480 volt

load center.

PBA [[]]

PS's report also documented that operators responded to the equipment

losses caused by the fire by initiating a transient of controlled reactor power reductions to

stabilize the plant at approximately 50 percent of rated power. The URI was opened

pending the

NRC staffs' characterization of this issue following review of

PBAPS's causal

evaluation and corrective actions.

PBAPS 's root cause report (

RCR) and the associated

IR 596616 for this event were reviewed to assess the identified issues. The

characterization of this issue as a finding and its risk significance are discussed below.

This

URI is closed. b.FindingsIntroduction. A Green self-revealing finding was identified for inadequate implementationof

WO instructions to verify the correct breaker frame size during the overhaul of a

compatible spare breaker for installation into the '4T4' 480 volt load center. This condition

resulted in a poor electrical connection between the primary disconnect fingers and the

switchgear bus stabs for one breaker in the '4T4' load center that ultimately resulted in a

fire that led to a plant transient and declaration of an Unusual Event (UE).Description. On February 27, 2007, operators reduced Unit 3 reactor power from 100percent to 50 percent RTP in response to the effects of a fire in the '4T4' 480 volt load

center.

RCR identified that a root cause for the configuration error was that standards,policies, and administrative controls (

RCR stated that the fire was caused by an electrical fault in one breaker

cubicle that occurred due to a poor electrical connection between the breaker primary

17Enclosuredisconnect fingers and the switchgear bus stabs. This poor electrical connection resultedfrom a configuration error that placed the wrong frame size breaker into the cubicle in the

'4T4' 480 volt load center creating a high resistance, high temperature connection. The

SPAC ) were not used. Specifically,

SPAC were notused, in that, the maintenance technicians did not strictly adhere to

WO instructions to

specifically verify the frame size during the overhaul of a spare breaker that was intended

to be placed into the breaker cubicle. The inspectors determined that this issue was a performance deficiency becausemaintenance technicians did not follow WO instructions to verify the correct breaker frame

size during the overhaul of a spare breaker. Analysis. This finding is greater than minor because it affected the human performanceattribute of the Initiating Event Cornerstone, in that, the incorrect frame size breaker was

installed in cubicle for which it was not sized. This mismatch caused an electrical fault

that led to a fire and a transient that upset plant stability. The inspectors evaluated the finding in accordance with

IMC 0609, Appendix A, "

SDP ofReactor Inspection Findings for At-Power Situations." The SDP Phase 1 screeningidentified that the finding was of very low safety significance (Green) because it did not

increase both the likelihood of a reactor scram and that mitigation equipment or functions

would not be available. The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance (work practices component) because maintenance technicians did

not follow WO instructions to specifically verify the frame size of a breaker during its

overhaul (IMC 0305 aspect H.4(b)). Enforcement. The inspectors determined that the finding did not represent a violation ofregulatory requirements because it involved the '4T4' 480 volt load center, a non-safety

related electrical bus. This finding will be tracked as FIN 05000278/2007003-01,

Inadequate Implementation of Work Order Instructions Caused the Installation of an

Incorrect Size Breaker and Resulted in a Fire in the '4T4' 480 Volt Load Center.2(Closed)

URI 05000277/2007002-05, Missed Procedure Step Resulted in UnplannedOverloading of the E-3

EDGURI 05000277/2007002-05 was opened in

NRC Inspection Report 050000277;05000278/2007002, pending the

NRC staffs' characterization of this issue following a

review of

PBA [[]]

PS's root cause analyses, corrective actions taken or planned, approved

procedures, and other documents. The characterization of this issue as a finding and its

risk significance are discussed below. This

URI is closed. b.FindingsIntroduction. A self-revealing (Green)

NCV of Technical Specification (TS) 5.4.1, wasidentified when operators inadequately implemented a surveillance test by missing a

18Enclosureprocedure step. The missed step placed the E-3 EDG in the isochronous mode ofoperation while it was synchronized to off-site power and resulted in an unexpected over-

loading of the E-3

EDG. Description. During the conduct of a E-3

EDG [[]]

ST on March 15, 2007, a licensed operatormissed the performance of a required step in a supporting system operating (

SO)

procedure. The omission of the procedure step placed the E-3 EDG in the isochronous

mode while synchronized with off-site power through a 4 kilovolt (kV) vital bus. This

condition resulted in unexpectedly loading the E-3 EDG beyond its 30-minute load rating.

The

ST -O-052-123-2, "E3 Diesel Generator

RHR Pump Reject Test," and the supporting

SO 52.A.1.B, "Diesel Generator Operations," directed the synchronization of the E-3

EDG,

in the droop mode, to a selected 4 kV bus to pick up the bus loads. The SO 52.A.1.B

procedure subsequently directed opening the off-site power feeder breaker to the 4 kV

vital bus (the missed step) before placing the EDG in the isochronous mode in

accordance with

ST -O-052-123-2. The inspectors reviewed

PBAPS 's root cause investigation report (IR 604364) tounderstand the underlying causes for this event. The inspectors noted that

PBA [[]]

PS

identified two root causes for this self-revealing event. First, the plant reactor operator

(PRO) did not adhere to the requirements of

HU -

AA-104-101, "Procedure Use and

Adherence" for "Level 1 - Continuous Use," procedures which requires that each

procedure step be read prior to being performed, performing each step in the sequence

specified, and signing off each step as complete prior to proceeding to the next step.

Specifically, procedure adherence broke down because the PRO allowed himself to be

distracted and lost his place in SO 52.A.1.B. Therefore, the off-site feeder breaker to the

E-33 bus was not opened in accordance with the

SO prior to transferring the E-3

EDG to

the isochronous load control mode per the

ST.T he second root cause for this event was inadequate supervisory oversight during acritical transition between the

ST and SO procedures. Specifically, the peer checker and

the control room supervisor were not directly observing the operation of the E-3 EDG at

the main control room panel during the critical transition between procedures. The

transition between procedures should have been identified as a critical step in the testing

evolution. This breakdown in crew teamwork resulted in the PRO performing a critical

step, without direct oversight, during an infrequently performed test of safety-related

equipment. As a result, no one challenged the

PRO 's decision to transfer the E-3

EDG to

the isochronous load control mode when system conditions did not support it.Based on the above, the inspectors determined that inadequately implementing asurveillance test by missing a procedure step was a performance deficiency. Analysis. The inspectors concluded the finding was more than minor because it wasassociated with the human performance attribute of the Mitigating Systems Cornerstone,

and impacted the cornerstone objective of ensuring the availability of E-3 EDG to respond

to initiating events, in that, after the EDG was overloaded, additional unavailability was

incurred to inspect the

EDG for damage before it was returned to service. The E-3

EDG

was inoperable for an additional 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months and was unavailable for an additional 12.5hours. Traditional enforcement does not apply since there were no actual safety

19Enclosureconsequences or potential for impacting the

NRC 's regulatory function, and the findingwas not the result of any willful violation of

NRC requirements. The inspectors completed a significance determination of this issue using

IMC 0609,"

SDP," Appendix A, "Determining the Significance of Reactor Inspection Findings for

At-Power Situations." The inspectors concluded that this finding affected the Mitigating

Systems Cornerstone and answered "No" to all relevant questions. Specifically, all other

EDG s remained operable and the actual loss of safety function for E-3

EDG was shorter

than its TS allowed outage time of seven days. Therefore, this finding was considered to

be of very low safety significance (Green).The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance (work practices component) because

PBA [[]]

PS personnel did not

follow procedures when the E-3 EDG was placed in the isochronous load control mode

with the E-3

EDG in parallel with the off-site power source. (

IMC 0305 aspect

H. 4(b))Enforcement.

TS 5.4.1 requires that procedures be implemented covering the activities inRegulatory Guide (RG) 1.33. RG 1.33, Appendix A, Section H.2.b requires that

surveillance procedures be developed for testing

EDG s. Applicable

ST-O-052-123-2,

Step 6.3.1, instructed the operators to synchronize and load the E-3 EDG to the 4 kV bus

being tested in accordance with

SO 52A.1.B. Step 4.4.16 of

SO 52A.1.B directed the

operators to open the off-site power source feeder breaker to the E-33 bus before placing

the

EDG controls in the isochronous load control mode. Contrary to the above, on March 15, 2007, operators missed

SO 52A.1.B, Step 4.4.16,and did not open the applicable off-site power breaker before returning to

ST -O-052-123-2, Step 6.3.2. Therefore, when the

PRO placed the E-3 EDG in the

isochronous load control mode in Step 6.3.2, there was an unexpected increase in E-3

EDG load and a trip of the E-3

EDG output breaker.PBAPS placed this issue in the

CAP by initiating

IR 604364. The corrective actions forthis event included: 1) the selective implementation of additional peer checking of

procedure performance place-keeping; and, 2) the E-3 EDG was inspected for potential

damage and tested before being returned to an operable status on March 17, 2007.

Because this violation was of very low safety significance (Green) and documented in

IR 604364, this finding is being treated as an

CAP as

NCV , consistent withSection

VI.A of the

NRC Enforcement Policy: NCV 05000277/2007003-02;05000278/2007003-02, Missed Procedure Step Resulted in Unplanned Overloading

of the E-3

EDG.. 3Personnel Performance - Failure of

DDFP a.Inspection ScopeThe inspectors reviewed corrective action documents listed in the Attachment to thisreport, and discussed the events surrounding the failure of the

DD [[]]

FP with the site fire

protection engineer. The inspectors reviewed Revisions 10 and 12 of ST-O-37D-340-2,

"DDFP Flow Rate Test," and Revision 2 of NOM-C-7.1, "Procedure Use."

20Enclosure b.FindingsIntroduction. A self-revealing Green

NCV was identified for failure to comply with

TS5.4.1, "Procedures," which required that procedures be established, implemented, and

maintained for the Fire Protection Program. Description.

PBAPS [[]]

TS 5.4.1.a, requires that procedures be established, implementedand maintained as recommended in Appendix A to RG 1.33, dated November 1972.

RG 1.33, Appendix A, Section 1, "Administrative Procedures," includes the fire protection

program. The Nuclear Operations Manual (NOM)-C-7.1, "Procedure Use," requires that

procedures be used for any task which has the potential to cause a system or component

to become inoperable.On May 23, 2007, during performance of

ST -O-37D-340-2, the

DDFP was declaredinoperable due to low discharge pressure. After running the

DD [[]]

FP, the procedure

directed cleaning of the cooling water strainer, but did not provide specific instructions on

how to perform this task. Without procedure guidance or instructions, operations

personnel performing the

DD [[]]

FP test closed an upstream hand valve to isolate the strainer

for cleaning. After reassembling the strainer, the operations personnel did not re-open the

hand valve. The cooling water was not properly realigned for service because equipment

manipulations were performed outside of procedure guidance. On May 24, 2007,

ST-O-37D-340-2 was re-performed with the cooling water supply isolated. The engine

was damaged during operation without cooling water as a result of the valve mis-

alignment. The

DDFP was subsequently returned to service on May 30, 2007, following repairs. Additionally, the

DDFP flow rate test procedure was revised to include specific instructions

for cleaning the cooling water strainer. The procedure was also revised to include

instructions for monitoring the engine cooling water and lubricating oil parameters during

engine operation. Based on the above, the inspectors determined that manipulating the

DDFP cooling watervalve without procedure guidance was a performance deficiency. Analysis. The inspectors concluded that the failure to use a procedure for cleaning andrestoring the

DDFP cooling water strainer was a more than minor finding because it was

associated with the degradation of a fire protection feature, in that, the

DD [[]]

FP was

rendered inoperable by damage to the engine. Traditional enforcement does not apply

since there were no actual safety consequences or potential for impacting the NRC's

regulatory function, and the finding was not the result of any willful violation of

NRC requirements. The inspectors assessed the finding using the Fire Protection

SDP (Appendix F to

IMC 0609) and determined the finding to be of very low safety significance (Green). The

finding was of low significance due to the motor-driven fire pump remaining operable

during the five days the

DD [[]]

FP was inoperable, and the small number of fire scenarios

which would impact the power supply to the motor-driven fire pump.

21EnclosureThe inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance (resources component) because procedure ST-O-37D-340-2 did not

provide complete and accurate instructions for cleaning the

DD [[]]

FP cooling water strainer.

(IMC 0305 aspect

H. 2©)Enforcement.

TS 5.4.1.a and NOM-C-7.1 require that procedures be used for equipmentmanipulations which could cause fire protection components to become inoperable.

Contrary to the above, procedures were not used when manipulating the

DD [[]]

FP cooling

water isolation valves on May 23, 2007, resulting in the

DD [[]]

FP being run on May 24, 2007,

without cooling water and sustaining engine damage. Because this failure to comply with

TS 5.4.1.a is of very low safety significance (Green) and has been entered into

IR 633532, this violation is being treated as an

CAP as

NCV , consistent with Section

VI.A of the

NRC Enforcement Policy:

NCV 05000277, 278/2007003-03, InadequateProcedure Adherence Results in Damage to the

DDFP.. 4(Closed) Licensee Event Report (

LER ) 05000277/2006002-00, AutomaticDepressurization System (ADS)

SRV DeficienciesOn September 28, 2006, engineering personnel determined that the 71B and 71G

SRVsdid not meet their allowable leak rate for the pneumatic actuation controls for the

ADS feature of the

SRV s. Additionally, the 71C

SRV, Serial Number 9S/N 83, did not properly

re-close on the fourth actuation during laboratory testing. The cause of the 71B and 71G

ADS [[]]

SRV pneumatic leakage is attributed to leakage from each of the SRV's actuator

diaphragm and solenoid valve. These leaks only occurred when the

SRV solenoid valveswere energized. The diaphragms and solenoid valves associated with the 71B and 71G

ADS [[]]

SRVs were replaced under work orders C0219044 and C0219034. As-left leak

testing was performed and the values were restored to an operable condition prior to plant

startup from the P2R16 Refueling Outage. A refurbished

SRV was installed in the 71C

SRV location to replace the S/N 83

SRV. The corrective actions to resolve the underlying

causes of this event are in the

CAP (

IR 539277). This licensee-identified violation was more than minor since it was associated with theEquipment Performance attribute of the Mitigating Systems Cornerstone and impacts the

cornerstone objective of ensuring the reliability, availability, and capability of systems that

respond to initiating events, in that, if the ADS system was called upon to actuate it's

operability would not be ensured. The inspectors evaluated this finding using IMC 0609,

Appendix A, "SDP of Reactor Inspector Findings for At-Power Situations," Phase 1

screening. Specifically, using the Mitigating Systems Cornerstone column, the inspectors

determined that a Phase 2 evaluation was required because the finding represented a

loss of system safety function. The inspectors concluded that the finding was of very low

safety significance (Green) because the success criteria for depressurization, on each of

the applicable worksheets, only required the use of 2 of 11 SRVs. A regional senior

reactor analyst reviewed and concurred with the inspectors risk assessment. This

licensee-identified finding involved a violation of TS 3.5.1, "Emergency Core Cooling

Systems." The enforcement aspects of this violation are discussed in Section 4OA7 of

this report. This LER is closed.

2Enclosure.5(Closed)

LER 05000277/2006004-00, Plant Modification Created Diesel GeneratorBuilding Carbon Dioxide Suppression Room Flooding VulnerabilityOn November 17, 2006, engineering personnel determined that a potential floodvulnerability had existed in the

EDG building carbon dioxide suppression room. A plant

modification performed in 1985 had installed a catch basin at the EDG building fuel oil

filling station, which is located outside the EDG building. The catch basin discharge was

tied into the EDG building's oily waste separator tank, upstream of the flood protection

isolation valve. This constituted an unanalyzed condition that degraded plant safety. In

the event of a design basis flood, a potential pathway existed for flood water to enter the

building through the floor drains. It was determined that the maximum credible flow rate

would have exceeded the capability of the floor drain sump and sump pumps. Under

design basis flood conditions, the

ESW system booster pumps and return valves, and the

HP [[]]

SW system return valves would be challenged to perform their safety function.

Corrective actions recommended for this issue were documented in IR 554800 and

included revision of the applicable special event procedure for floods to mitigate this

condition. This finding is more than minor because it was associated with a degraded condition thatcould concurrently influence mitigation equipment. Specifically, with the degraded flood

barrier for the

EDG building carbon dioxide suppression room, the

ESW system booster

pumps and return valves and the

HP [[]]

SW system return valves would be challenged to

perform their safety function under design basis flood conditions. The

NRC [[]]

IMC 0609,

Appendix G, "Shutdown Operations SDP," applies because the plant would be shutdown,

at 112', in accordance with plant procedures, before flooding of EDG building would begin

to occur at the 128' elevation, as noted in the

LER. Also, as noted in the

LER, the design

basis flood would be expected to reach the 132' elevation. A Phase 1 SDP was

performed using Checklist 5 of IMC 0609, Appendix G, Attachment 1. The inspectors

determined that a Phase 2 or 3 SDP was required because the finding:*Increased the likelihood that a loss of decay heat removal will occur due to afailure of its support systems; *Would degrade the ability to cope with a loss of offsite power; and

Would degrade the ability to establish an alternate core cooling path if decay heatremoval cannot be re-established for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The inspectors determined that the finding was of very low safety significance (Green)because of: the very low likelihood of occurrence of a design basis flood reaching the

2' elevation; flood alarms in the EDG building carbon dioxide suppression room that

would enable operators to take actions to stop the flooding; or operators could manually

operate the service water system return valves. A regional senior resident analyst

reviewed and concurred with the inspectors risk assessment. This licensee-identified

finding regarding the installation of a modification that placed the station in an unanalyzed

condition involved a violation of 10 CFR 50.59. The enforcement aspects of this violation

are discussed in Section 4OA7 of this report.

23Enclosure4OA5Other ActivitiesAs a plant status activity, the inspectors used guidance in

NRC [[]]

IP 60855.1, "Operation ofan Independent Spent Fuel Storage Installation at Operating Plants," to selectively verify

that

PBA [[]]

PS performed dry cask loading in a safe manner and in compliance with

approved procedures and work order instructions.4OA6Meetings, Including Exit.1Exit Meeting SummaryOn July 20, 2007, the resident inspectors presented the inspection results to Mr.

J. Grimes and other

PBAPS staff, who acknowledged the findings. The inspectors

asked the licensee whether any of the material examined during the inspection should be

considered proprietary. No proprietary information was identified. .2Annual Assessment MeetingOn April 4, 2007, Mr. Paul Krohn, Mr. Mel Gray, the resident inspection staff, and otherNRC staff held a public meeting with Mr. Joe Grimes and other

PBA [[]]

PS staff, to discuss

the results of the

NRC 's assessment of performance at

PBAPS for the period January 1,

2006 through December 31, 2006. The handouts from the meeting are available

electronically from the

NRC 's document system (

ADAMS ) under accession number

ML 071000066. Following the meeting, the

NRC staff held a session to accept public

comments and respond to public questions.4OA7Licensee-Identified ViolationsThe following violations of very low safety significance (Green) were identified by thelicensee and are violations of

NRC requirements which meet the criteria of Section

VI of

the

NRC Enforcement Policy,

NUREG -1600, for being dispositioned as

NCV s.*10

CFR 50.54(q) requires that the licensee shall follow and maintain in effectemergency plans which meet the standards in 50.47(b) and the requirements in

Appendix

E. The Exelon Nuclear Standardized Radiological Emergency Plan forPeach Bottom, Part

II, Section E.2 b.1 states for State/Local Agencies: A

notification shall be made within fifteen (15) minutes of the initial emergency

classification. Contrary to this, on February 27, 2007, during an emergency event,

Peach Bottom personnel failed to notify one local county within 15 minutes of an

initial emergency declaration (Unusual Event); the notifications were completed in

minutes. The notification was not made in a timely manner because the

primary phone link to the county was not available. Plant procedures require the

notifications to be made using a backup phone. This finding is of very low safety

significance (Green) because the notification was late by only 3 minutes, backup

communication equipment was available, and procedures were available to use

the backup communication equipment. This was entered in

CAP as IR

596641.

24Enclosure*10 CFR 50.59, "Changes, Tests, and Experiments," requires, in part, that thelicensee may make changes in the facility as described in the safety analysis

report without prior Commission approval, unless the proposed change involves a

change in the TSs incorporated in the license or an unreviewed safety question

(USQ). A proposed change shall be deemed to involve a USQ, in part, if the

consequences of an accident or a malfunction of equipment important to safety

previously evaluated in the safety analysis report may be increased. Contrary to

this, in 1985, a change to the facility was made that remained in place until

November 2006, without analyzing whether a USQ existed. Specifically, as

documented in Section 4OA3.5, a plant modification performed in 1985 introduced

a potential flood vulnerability for the EDG building carbon dioxide suppression

room. The flood vulnerability posed by this change constituted an unanalyzed

condition that degraded plant safety. This was identified in

CAP as

IR 554800. This finding is of very low safety significance (Green) because the

likelihood of a design basis flood that could affect mitigation equipment is very

small and manual operator action could be taken to mitigate the effects of a design

basis flood.*TS 3.5.1, "Emergency Core Cooling Systems," requires that the

ADS function offive

SRVs be operable. TS 3.5.1, Action H, requires the plant to be brought to

Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if two or more SRVs are inoperable. Contrary to the above, on

September 28, 2006, the pneumatic actuation controls for the ADS function of two

SRVs (71B and 71G) did not meet their allowable leak rate acceptance criteria.

Specifically, the as-found leak rates for the 71B and 71G SRVs were documented

as off-scale and were in excess of the allowable the leak rate limit of 100 cc/min.

Unit 2 was shutdown and in a refueling outage when the event was discovered.

However, Unit 2 had been operating for the previous 367 days. This issue was

entered in

IR 539277. As documented in Section 4

CAP as

OA3.4, a

Phase 2 SDP determined that the finding was of very low safety significance

(Green) because the success criteria for depressurization, on each of the

applicable

SDP notebook worksheets, only required the use of 2 of 11

SRV s.ATTACHMENT:

SUPPLE [[]]

MENTAL [[]]

INFORM [[]]

ATION A-1AttachmentSUPPLEMENTAL

INFORM [[]]

ATIONK EY

POINTS [[]]

OF [[]]

CONTAC [[]]

TExelon Generation Company PersonnelJ. Grimes, Site Vice PresidentM. Massaro, Plant Manager

N. Alexakos, Manager, Engineering-Programs

J. Armstrong, Regulatory Assurance Manager

C. Behrend, Engineering Director

G. Jardel, Manager, Emergency Preparedness

C. Jordan, Chemistry Manager

D. Lewis, Operations Director

H. McCrory, Radiation Protection Technical Support Manager

M. Ross, Radwaste, Environmental Supervisor

G. Stathes, Maintenance Director

S. Taylor, Manager, Radiation Protection

T. Van Wyen, Operations Training Manager

A. Wasong, Training Director

NRC PersonnelF. Bower, Senior Resident InspectorM. Brown, Resident Inspector

R. Fuhrmeister, Senior Project Engineer

R. Nimitz, Senior Health Physicist

N. Perry, Sr. Emergency Response Coordinator

R. Cureton, Emergency Preparedness Inspector

LIST [[]]

OF [[]]

ITEMS [[]]

OPENED ,

CLOSED ,

AND [[]]

DISCUSSEDOpenedNone.Opened and Closed05000278/2007003-01FINInadequate Implementation of WOInstructions Caused the Installation of

an Incorrect Size Breaker and

Resulted in a Fire in the '4T4' 480 Volt

Load Center (Section

4OA 3.1)05000277, 278/2007003-02

NCVM issed Procedure Step Resulted inUnplanned Overloading of the E-3

EDG (Section 4

OA3.2)

A-2Attachment05000277, 278/2007003-03NCVInadequate Procedure AdherenceResults in Damage to the

DD [[]]

FP (Section

4OA 3.3)Closed05000277/2007002-04

URIIncorrect Size Breaker Resulted in aFire in the '4T4' 480 Volt Load Center

(Section

4OA 3.1)05000277/2007002-05

URI Missed Procedure Step Resulted inUnplanned Overloading of the E-3

EDG (Section 4

OA 3.2)05000277/2006002-00LERADS

SRV Deficiencies(Section 4

OA 3.4)05000277/2006004-00

LE [[]]

RPlant Modification Created DieselGenerator Building Carbon Dioxide

Suppression Room Flooding

Vulnerability (Section

4OA 3.5)DiscussedNone.

LIST [[]]

OF [[]]

DOCUME NTS

REVIEW [[]]

EDS ection 1R01: Adverse WeatherWC-AA-107, Revision 4, Seasonal ReadinessOP-AA-108-111-1001, Revision 2, Severe Weather and Natural Disaster Guidelines

PB -108-111-1001, Revision 3, Preparation for Severe Weather

RT -O-040-610-2, Revision 12, Outbuilding

HVAC and Equipment Inspection for SummerOperationSO

52A. 1.B, Revision 39, Diesel Generator OperationsSection 1R04: Equipment Alignment

COL [[]]

52A. 1.A-3, Revision 12, E-3 Diesel Generator Normal Standby

SO 53.7.A - App 1, Revision 0, Removal of 220-08 Line from Service

COL 13.1.A-2, Revision 19,

RCIC System

COL 33.1.A-2, Revision 20,

ESW System (Unit 2 and Common)

COL 32.1.A-2, Revision 10,

HPSW System

SO 32.1.A-2, Revision 12,

HPSW System Startup and Normal Operatoins

A-3AttachmentP&ID DiagramM-315 Sheet 1, Revision 64,

ESW and

HPSW SystemsM-315 Sheet 4, Revision 53,

ESW and

HPSW Systems

M-315, Sheet 1, Revision 65,

ESW and

HPSWSystemsM-330, Sheet 1, Revision 35, Emergency Cooling System

M-361, Sheet1, Revision 80, RHR System

M-361, Sheet 2, Revision 67,

RHR SystemSection 1R05: Fire Protection

PF -63, Revision 1, Prefire Strategy Plan Unit 3 Reactor Bldg.

RCIC Room, 88' Elevation

PF -70, Revision 2, Prefire Strategy Plan Standby Gas Treatment Room, Radwaste Building,91' 6" ElevationPF-13H, Revision 3, Prefire Strategy Plan North

CRS Equipment and West Corridor, Unit 3Reactor Building, 135' Elevation

PF -55, Revision 3, Prefire Strategy Plan, Fire Zone 55, Unit 3 Refuel Floor, Reactor Building,234' ElevationPF-13D, Revision 1, Prefire Strategy Plan 3 'A' & 3 'C' Core Spray Rooms, Reactor Building,91'6" Elevation, Fire Zones 13D &

13EPF -60, Revision 1, Prefire Strategy Plan, Unit 2 Reactor Building

RCIC Room, 88' Elevation

PF -127, Revision 4, Prefire Strategy Plan, Unit 2 Emergency Battery/ Switchgear Room andRadwaste Corridor,

TB -135PF-132, Revision 4, Prefire Strategy Plan, Diesel Generator Building, Elevation 127', FireZone 132PF-151, Revision 3, Prefire Strategy Plan, Unit 2 Main Transformer Yard, Fire Zone 151

PF -164, Revision 0, Prefire Strategy Plan, 2 Startup Switchgear Building, Fire Zone 164Section 1R12: Maintenance Effectiveness

IR 607398, Functional Failure of

3AE 015 During '4T4' Breaker Fire

IR 596616, Fault at Unit 3 'B' Iso-Phase Cooler Fan Breaker in 4T4

IR 614945, Potential Extent of Condition Concern for

MCC Bucket Stabs

IR 619579, 480 V Breaker Interference Angle Location Incorrect

IR 617890, Conflicting Data on Cubicle Size of 2 'A'

EHC Pump Breaker

IR 599184, Extent of Condition Walkdown of Unit 2 480 V Load Center Bus

IR 606397, Perform

ITE Rejection Tab Walkdown

IR 599203, Extent of Condition Walkdown of Unit 3 480 V Load Center Bus

IR 599208, Extent of Condition Walkdown of Common 480 V Load Center Bus

IR 634973,

IR 634971,

IR 634962,

IR 634964,

IR 634966,

IR 600797, 2007 Buried Pipe Program Inspections

IR 634965,

ITE Breaker Found With No Rejection Tab

IR 623638,

EOC Generate

PM per

PCM Template Requirements

IR 623646,

EOC Generate

PM per

PCM Template Requirements

A-4AttachmentIR 623635,

EOC Generate

PM per

PCM Template Requirements

IR 603279, Inspect and Clean

ESW X-Tie Piping (

HV -512A-B) WW 0730

IR 632688, 2 'A'

EHC PP Breaker Cubicle Frame Size Incorrect

IR 589654, Potential For Silt Buildup in the

ESW Pump Crosstie Piping

ACPS 07-0-002,

HV -0-33-512A, A ESW Pump Discharge Loop X-tie

ST -O-033-300-2, Revision 31,

ESW , Valve, Unit Cooler and ECT Functional Inservice Test

ACPS 07-0-002,

HV -0-33-512A, A ESW Pump Discharge Loop X-tie

ST -O-033-300-2, Revision 31,

ESW, Valve, Unit Cooler and ECT Functional Inservice Test

Performance Monitoring - Unavailability - System 33 (ESW) - Jun 2005 -> Jun 2007

Clearance 07000529, Emergency Cooling Water Pump Discharge Valve

ER -

AA -5400, Revision 0, Buried Piping and Raw Water Corrosion Program Guide

ER -

AA -5300, Revision 0, Raw Water Corrosion Program Guide

ER -

AA-5400-1002, Revision 0, Buried Piping Examination GuideSection 1R13: Maintenance Risk Assessments and Emergent Work ControlWC-AA-101, "On-line Work Control Process"Adverse Condition Monitoring and Contingency Plan (CAMP), 3 'A' Recirculation Pump Seal

Unstable Second Stage Seal Temperature and Increasing Second Stage Seal Pressure, Dated 04/17/2007AR A1612541, Rising 3 'A' Recirculation Pump #2 Seal Temperature

AR A1610537, High Lube Oil Temperature Alarm During E-2

EDG Run

AR A1613094-01, Technical Evaluation:

CRD Suction Source Swap from Condensate to Unit

3 CST [[]]

IR 623723, Bolt and Heli-coil Found Damaged at Disassembly on 00T634

SF-220, Revision 21, Spent Fuel Cask Loading and Transport Operations

A1406063, Review of Mod 79-028 Recirculation Seal Pressure Bleed Off

EC 360901, Exelon Fleet Reactor Recirculation Pump Seal Condition Monitoring Template

IR 620785, Continuous Venting of the Recirculation Seals not EvaluatedAO

2A. 16-3, Revision 2, Manual Adjustment of Recirculation Pump Seal Second Stage Pressure

SO 2A.1.C-3, Revision 10, Operation of the Recirculation Pump Seal Purge System

A1439223,

3AP 034: Seal Hi Temp Alarm & Hi 2nd Stage Pressure

ACMP - Unit 3, 3 'B' Recirculation Pump Seal Increasing Second Stage Seal Pressure

A1613202, 3 'B' Recirculation Pump 2nd Stage Seal PressureIR 619609, 3 'B' Recirculation Pump 2nd Stage Seal PressureARC 30C204M A-1, Revision 4 - A Recirculation Pump Seal Stage 2 Hi Flow

ARC 30C204M A-2, Revision 6 - A Recirculation Pump Seal Stage 2 Lo Flow

PB -108-101-1002, "Guidelines for Control of Protected Equipment," Revision 4

WC -

AA-101, "On-Line Work Control Process," Revision 13

IR 626534, Equipment Not Protected as Required.

IR 624653, Protected Equipment List for 2

SUE Outage Incomplete

IR 617946, Protected Equipment List Issued 4/16/07 Initially Incomplete

IR 504032, Exaggerated Paragon List of Protected Equipment

IR 462364-18-04, Paragon Refresher Training

IR 624599, U3

RHR Pump Testing Not Performed per Schedule

IR 634657,

PRA Support for Protecting Equipment

IR 474569-17-08, Develop a Tutorial that Help Crews with Paragon

IR 624599, U3

RHR Pump Testing Not Performed per Schedule

IR 644648, Inadequate Guidance in

WC-AA-101 for Protecting Equipment

A-5AttachmentARC-216 20C212L D-1, Revision 5, C Air Comp TroubleSO 36A.7.A-2, Revision 3, Unit 2 'C' Air Compressor Shutdown

ON-119, Revision 14, Loss of Instrument Air

ARC-216 20C212L D-2, Revision 2 Service Air Header Lo Press

ARC-316 20C212L D-2, Revision 1, Service Air Header Lo Press

R1032642,

3CK 001 -

PM Perform Annual PM on Compressor

SO 36A.1.A-2, Revision 2, Unit 2 'C' Air Compressor Return-to-Service and Service Air Systems Return to Normal Operation

IR 642127, Critique

IR on Loss of Service Air to Unit 2 and Unit 3DrawingsP&I Diagram M-356,

CRD Rod Drive Hydraulic System Part A, Sheet

2P&I Diagram M-353, Reactor Recirculation Pump SystemSection 1R15: Operability Evaluations

IR 615433, E-4

EDG - 10

CFR [[21 Notification for Cam Roller Bushing Material Issue Fairbanks Morse Engine Notification Report Serial Number 06-04, 10 CFR 21 Notification, CamRoller Bushing Incorrect Material, dated April 9, 2007Event Notification Number 43294, Part 21 Notification - Diesel Cam Roller Bushing Failures]]

IR 388397-04, Prompt Investigation of 3 'A'

RRP #2 Seal Cavity Temperature HighAdverse Condition Monitoring and Contingency Plan (CAMP), 3 'A' RRP Unstable Second

Stage Seal Temperature and Increasing Second Stage Seal Pressure, dated 04/17/2007Operational and Technical Decision Making (OTDM) No. 07-01, 3 'A'

RRP Seal Issues, dated04/17/2007Abnormal Operations (

AO ) procedure

2A. 16-3, Manual Adjustment Recirculation Pump SealSecond Stage Pressure

OTDM No. 07-02, 3 'A'

RRP Seal Temperatures - Re-align

CRD Suction Source fromCondensate System to U3

CST , dated 04/20/2007

AR A1613094-01, Technical Evaluation:

CRD Suction Source Swap from Condensate to Unit 3

CST [[]]

PBAPS Technical Requirements Log, Item Number 07-3-080,

PTRM 3.6, Function 7, MainSteam Relief Valves, dated May 17, 2007Adverse Condition Monitoring Plan:

DPT -2-02-117

DH Sensing Line Leakage, datedMay 24, 2007A1615458, Small Leak on DPT-2-02-117D Line Snubber Threaded Cap

C0221439, Replace Snubber During an Outage

PB [[]]

ECR 03-00326 000, Revise Instrument Rack Drawings with a Note for SnubbersSection 1R19: Post-Maintenance TestingAR A1610537, High Lube Oil Temperature Alarm During E-2

EDG RunR1049367, Unit 3

HCU 50-43: HCU Overhaul

ST -R-003-480-3, Average Scram Times for

ODYN /B Minimum Critical Power Ratio (MCPR) RequirementsC0216504, PS-2-13-067-01: Replace Pressure Switch

ST -O-013-301-2, Revision 31,

RCIC Pump, Valve, Flow and Unit Cooler Functional andInservice Test, Conducted on April 5, 2007

A-6AttachmentC0215740,

2BG 002, Replace Endbell

MA -AA-716-230-1002, Revision 1, Vibration Analysis/ Acceptance Guideline

MA -

AA -716-230-1003, Revision 1, Thermography Program Guide

SO 60F.1.A-2, Revision 9, Reactor Protection System

MG Set and Power Distribution System Startup from Dead Bus ConditionR0629147, 3R4-U-C (7033B), Perform MCU Inspection

A1619582,

3CP 343: Pump/Motor Found Seized during Breaker

PMT [[]]

IR 638369, 3C Glycol Pump found seized during breaker

PMT [[]]

SO 8G.6.A-3, Revision 3, Placing a Standby Off-Gas Glycol Pump in Service and Placing the InService Pump in Standby or OffSection 1R22: Surveillance TestingS12T-

MIS -8547-C1CQ, Revision 13, Calibration/Functional Check of Channel C Group 1, 4 and 5 of

PCIV Logic for

TSS -80547CST-R-003-485-3,

CRD Scram Insertion Timing of Selected Control Rods, Revision 19, completed May 5, 1997Section 1R23: Temporary Plant Modifications

IR 618478, Provide Supplemental Cooling to the 3 'A'

RR Seal Purge Line

IR 625092, Equipment Discovered on Floor Hatch H11 in Unit 3 Reactor Building

WO C0221034,

TCCP 07-00172, Install Cooling Unit

AR A1613094, Provide Supplemental Cooling to the 3 'A'

RR Seal Purge Line

SP [[]]

SO. 005-3, Revision 1, Routine Inspection of the 3 'A' Recirculation Seal Purge SupplementalCooling SystemSection

1EP 2: Alert and Notification System (

ANS ) EvaluationPeach Bottom Nuclear Power Plant Upgraded Public Alert and Notification Report, April

2005FEMA [[]]

ANS Design Report, December

2005EP -

MA -121-1002 "Exelon East Alert and Notification System (ANS) Program," Revision 4

MA -121-1004 "Exelon East ANS Corrective Maintenance," Revision 4

MA -121-1005 "Exelon East ANS Preventive Maintenance Program," Revision 3

MA -121-1006 "Exelon East

ANS Siren Monitoring, Troubleshooting, and Testing," Revision 5Corrective Maintenance Field Work Instructions for

ANS Control Points, Repeaters and Sirens, Approved December 2004Preventative Maintenance Field Work Instructions for

ANS Control Points, Repeaters and Sirens

IRs:0043349400565056

00352078

005970650048176300451040

00520830

005213210053315700541478

00596641

A-7AttachmentSection

1EP 4: Emergency Action Level (

EAL ) and Emergency Plan ChangesEP-AA-120-1001 "10

CFR 50.54(q) Change Evaluation," Revision 406-12 "

ERO Training and Qualification"

TQ -

AA-113, Revision 7

06-16 "Radiological Emergency Plan"

AA-1000, Revision 17

06-33 "EP Plan Administration"

AA-120, Revision 7

06-96 "Emergency Preparedness Advisory Committee"

AA-120-1004, Revision 0

06-97 "Quarterly Satellite Phone Test"

MA-124-1004, Revision 0

06-99 "EP Fundamentals"

ANS Corrective Maintenance Program"

AA-1101, Revision 3

06-101"Exelon East

ANS Program"

EP-MA-121-1002, Revision 4

06-102"Exelon East

EP-MA-121-1004, Revision 4

06-103"Exelon East

ANS Preventative Maintenance Program"

EP-MA-121-1005, Revision 3

06-108"ERO Fundamentals"