IR 05000277/2007003
| ML072120599 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/30/2007 |
| 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 | |
| IR-07-003 | |
| Download: ML072120599 (22) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I475 ALLENDALE ROADKING OF PRUSSIA, PENNSYLVANIA 19406-1415 July 30, 2007 Mr. Christopher President and CNO Exelon NuclearExelon Generation Company, LLC200 Exelon Way Kennett Square, PA 19348
SUBJECT: PEACH BOTTOM ATOMIC POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000277/2007003 AND 05000278/2007003
Dear Mr. Crane:
On June 30, 2007, the United States Nuclear Regulatory Commission (NRC) completed aninspection at your Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The enclosedintegrated inspection report documents the inspection results, which were discussed on July 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 three self-revealing findings of very low safety significance (Green). Two ofthese findings were determined to involve violations of NRC requirements. Additionally, threelicensee-identified violations which were determined to be of very low safety significance are listedin 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-citedviolations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contestany NCV in this report, you should provide a response within 30 days of the date of this inspectionreport, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: DocumentControl 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.
C. M. Crane2In accordance with 10 Code of Federal Regulations (CFR) 2.390 of the NRC's "Rules of Practice,"a copy of this letter, its enclosures, and your response (if any) will be available electronically forpublic inspection in the NRC Public Document Room or from the Publicly Available Records(PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from theNRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic ReadingRoom).
Sincerely,/RA/Paul G. Krohn, Chief Reactor Projects Branch 4Division of Reactor ProjectsDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56
Enclosures:
Inspection Report 05000277/2007003 and 05000278/2007003
w/Attachment:
Supplemental Informationcc w/encl:Chief Operating Officer, Exelon Generation Company, LLC Site Vice President, Peach Bottom Atomic Power Station Plant Manager, Peach Bottom Atomic Power Station Regulatory Assurance Manager - Peach Bottom Manager, Financial Control & Co-Owner Affairs Vice President, Licensing and Regulatory Affairs Senior Vice President, Mid-Atlantic Senior Vice President - Operations Support Director, Licensing and Regulatory Affairs J. Bradley Fewell, Assistant General Counsel, Exelon Nuclear Manager Licensing, PBAPS Director, Training Correspondence Control Desk Director, Bureau of Radiation Protection, Department of Environmental Protection R. McLean, Power Plant and Environmental Review Division (MD)
G. Aburn, Maryland Department of Environment T. Snyder, Director, Air and Radiation Management Administration, MD Department of the Environment Public Service Commission of Maryland, Engineering Division Board of Supervisors, Peach Bottom Township B. Ruth, Council Administrator of Harford County Council Mr. & Mrs. Dennis Hiebert, Peach Bottom Alliance TMI - Alert (TMIA)
J. Johnsrud, National Energy Committee, Sierra Club Mr. & Mrs. Kip Adams E. Epstein, TMI Alert R. Fletcher, Department of Environment, Radiological Health Program C. M. Crane2In accordance with 10 Code of Federal Regulations (CFR) 2.390 of the NRC's "Rules of Practice," a copy ofthis letter, its enclosures, 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 NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/r eading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/Paul G. Krohn, Chief Reactor Projects Branch 4 Division of Reactor ProjectsDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56
Enclosures:
Inspection Report 05000277/2007003 and 05000278/2007003
w/Attachment:
Supplemental InformationDistribution w/encl:S. Collins, RAR. Fuhrmeister, DRP M. Dapas, DRAT. Setzer, DRP J. Lamb, RI OEDO F. Bower, DRP - Senior Resident Inspector H. Chernoff, NRRM. Brown, DRP - Resident Inspector J. Hughey NRR, PMS. Schmitt - Resident OA P. Bamford, PM, BackupRegion I Docket Room (with concurrences)
J. Lubinski, NRRROPreports@nrc.govP. Krohn, DRP SUNSI Review Complete: __PGK_______ (Reviewer's Initials)DOCUMENT NAME: C:\FileNet\ML072120599.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 copyML072120599OFFICERI/DRP Rl/DRP NAMEFbower/PGK forPKrohn/PGKDATE07/24/0707/24/07OFFICIAL RECORD COPY U. S. NUCLEAR REGULATORY COMMISSIONREGION IDocket Nos.:50-277, 50-278 License Nos.:DPR-44, DPR-56 Report No.:05000277/2007003 and 05000278/2007003 Licensee:Exelon Generation Company, LLC Facility:Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Location:Delta, Pennsylvania Dates:April 1, 2007 through June 30, 2007 Inspectors:F. 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 InspectorApproved by:Paul G. Krohn, ChiefReactor Projects Branch 4 Division of Reactor Projects ii
SUMMARY OF FINDINGS
IR 05000277/2007-003, 05000278/2007-003; 04/01/2007 - 06/30/2007; Peach Bottom AtomicPower Station (PBAPS), Units 2 and 3; Event Followup. The report covered a 3-month period of inspection by resident inspectors, a senior projectengineer, and announced inspections by a senior health physicist, a senior emergency response coordinator, and an emergency preparedness inspector. Three Green findings, two of which were NCVs, were identified. The significance of most findings is indicated by their color (Green,
White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006. A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
A self-revealing finding was identified for inadequate implementation ofwork order (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).This finding is greater than minor because it affected the human performanceattribute of the Initiating Event Cornerstone, in that, an incorrect frame size breaker was installed into a cubicle for which it was not sized. This mismatch caused an electrical fault that led to a fire and a plant transient that upset plant stability. 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 of human 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)). (Section 4OA3.1)*Green. A self-revealing NCV of Technical Specification (TS) 5.4.1, was identifiedwhen operators inadequately implemented a surveillance test by missing a procedure step. The missed step placed the E-3 emergency diesel generator (EDG) in the isochronous mode of operation while it was synchronized to off-site power and resulted in an unexpected over-loading of the E-3 EDG. This finding is more than minor because it was associated with the humanperformance attribute of the Mitigating Systems Cornerstone, and impacted the cornerstone objective of ensuring the availability of the E-3 EDG to respond to initiating events. This finding is of very low safety significance (Green) because all other EDGs remained operable and the actual loss of safety function of the E-3 EDG was less than the TS allowed outage time of seven days. The inspectors ivdetermined that this finding had a cross-cutting aspect in the area of humanperformance (work practices component) because PBAPS 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)). (Section 4OA3.2)*Green. A self-revealing NCV of TS 5.4.1, was identified when operatorsmanipulated a diesel-driven fire pump (DDFP) cooling water valve outside of procedure guidance. The improper manipulation led to misalignment of the DDFP cooling water that subsequently damaged the engine during operations without cooling water. The failure to use a procedure for cleaning and restoring the DDFP cooling waterstrainer was a more than minor finding because it was associated with the degradation of a fire protection feature, in that, the DDFP was rendered inoperable by damage to the engine. Using the Fire Protection SDP, the finding was determined to be of very low safety significance due to the motor-driven fire pump remaining operable during the five days the DDFP was inoperable, and the small number of fire scenarios which would impact the power supply to the motor-driven fire pump. This finding had a cross-cutting aspect in the area of human performance (resources component) because procedure ST-O-37D-340-2 did not provide complete and accurate instructions for cleaning the DDFP cooling water strainer (IMC 0305 aspect H.2©). (Section 4OA3.3)
B.Licensee-Identified Violations
Three violations of very low safety significance (Green), that were identified by thelicensee, have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's CAP. The violations and corrective actions are listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant StatusUnit 2 began the inspection period at 100 percent full rated thermal power (RTP) until April 27, 2007, when power was reduced to 58 percent for planned waterbox cleaning, control rod testing, 2 'A' reactor feed pump (RFP) maintenance, and other planned maintenance and testing. On April 28, 2007, the unit returned to full power where it remained until the end of the inspection period, except for brief periods to support planned testing and rod pattern adjustments. Unit 3 began the inspection period at 100 percent full RTP until April 16, 2007, when anunplanned power reduction to 84 percent was performed in response to rapidly increasing 3 'A' reactor recirculation pump shaft seal temperatures. The unit returned to full power on April 18, 2007. On May 4, 2007, power was reduced to 59 percent for planned waterbox cleaning, control rod testing, and 3 'C' RFP maintenance. The unit returned to full power on May 5, 2007. On May 11, 2007, power was reduced to 65 percent for a planned control rod pattern adjustment and RFP testing, and the unit returned to full power on May 12, 2007. On June 15, 2007, power was reduced to 82 percent for a rod pattern adjustment and planned maintenance on a feedwater heater drain line. The unit was returned to full power on June 16, 2007, where it remained until the end of the inspection period, except for brief periods to support planned testing and rod pattern adjustments.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 System Sample; 1 Site Sample).1Summer Seasonal Readiness
a. Inspection Scope
The inspectors performed one seasonal readiness sample that included a review of threeventilation systems. Specifically, the inspectors reviewed the procedures listed in 1 to the report, and verified summer ventilation system alignment for the diesel generator building, circulating water pump structure, and circulating water pump screen house.
b. Findings
No findings of significance were identified.
2Enclosure.2Adverse Weather Event Review
a. Inspection Scope
On June 13, 2007, a tornado warning was issued for an adjacent county. The inspectorsreviewed PBAPS's actions taken to respond to potential adverse environmental conditions from severe thunderstorms that entered the area. High winds, lightning, rain, and reports of hail were experienced at the site. The inspectors observed that PBAPS's personnel consulted procedure OP-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. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04Q - 3 Partial Walkdown Samples).1Partial Walkdown
a. Inspection Scope
The 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 PBAPS 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' ESW Out-of-Service.
b. Findings
No findings of significance were identified..2Complete System Walkdown (71111.04S - 1 Sample)
a. Inspection Scope
During the week of June 25, 2007, the inspectors performed one complete Unit 2 highpressure service water (HPSW) system walkdown of the accessible portions of the 3Enclosuresystem. The full walkdown was performed to identify any discrepancies which couldimpact the Unit 2 HPSW 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. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05Q - 10 Samples)Fire 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 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), RCIC 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 RCIC (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. Findings
No findings of significance were identified.
4Enclosure1R11Licensed Operator Requalification Program (71111.11Q - 1 Sample) Resident Inspector Quarterly Review
a. Inspection Scope
On June 12, 2006, the inspectors observed operators in PBAPS'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. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12Q - 2 Samples)
a. Inspection Scope
The 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 PBAPS'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. Findings
No findings of significance were identified.
5Enclosure1R13Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 Samples)
a. Inspection Scope
The 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 PBAPS'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 2 nd 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. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15 - 5 Samples)
a. Inspection Scope
The 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. 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 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 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. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22 - 5 Samples) [3 Routine Samples; 1 IST Sample; 1Reactor Coolant System (RCS) Leakage Sample]
a. Inspection Scope
The 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 TSs-80547C; and*ST-R-003-485-3, CRD Scram Insertion Timing of Selected Control Rods.b.FindingsNo findings of significance were identified.
1R23 Temporary Plant Modifications (71111.23 -1 Sample)
a. Inspection Scope
The 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. Findings
No findings of significance were identified.Cornerstone: Emergency Preparedness1EP2Alert 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 PBAPS'sANS. During this inspection, the inspectors interviewed emergency preparedness (EP)staff responsible for implementation of the ANS testing and maintenance. 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.
b. Findings
No 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 IRs 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, 10 CFR 50.47(b)(2) and related requirements of 10 CFR 50, Appendix E were used as reference criteria.
b. Findings
No findings of significance were identified. 1EP4Emergency Action Level (EAL) and Emergency Plan Changes (71114.04 - 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 10 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. Findings
No 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. Findings
No findings of significance were identified.1EP6Drill Evaluation (71114.06 - 1 Sample)Off-Year Exercise (Drill)
a. Inspection Scope
The 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 PBAPS'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 PBAPS's identified issues against the inspectors' observations to determine whether PBAPS 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. Findings
No findings of significance were identified.2.RADIATION SAFETYCornerstone: Public Radiation Safety 2PS2Radioactive Material Processing and Transportation (71122.02 - 5 Samples).1Inspection Planning/In-Office Inspection
a. Inspection Scope
The inspectors reviewed the solid waste system description in the UFSAR 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 2PS2.6)
b. Findings
No findings of significance were identified.
.2 Radioactive Waste System Walkdown
a. Inspection Scope
The 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 UFSAR; 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. Findings
No findings of significance were identified..3Waste Characterization and Classification
a. Inspection Scope
The 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 10 CFR 20, 10 CFR 61, 10 CFR 71, the UFSAR, the PCP, applicable NRC BTPs, and Exelon procedures.
b. Findings
No findings of significance were identified.
12Enclosure.4Shipment Preparation
a. Inspection Scope
The 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. Findings
No findings of significance were identified.
.5 Shipment Records and Documentation
a. Inspection Scope
The 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. Findings
No findings of significance were identified..6 Audits and Assessments of Radioactive Waste Handling
a. Inspection Scope
The 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, ASSA-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. Findings
No findings of significance were identified.4.OTHER ACTIVITIESCornerstones: Barrier Integrity & Emergency Preparedness 4OA1Performance Indicator (PI) Verification (71151 - 7 Samples)
.1 Barrier Integrity PIs ( 71151 - 4 Samples)
a. Inspection Scope
The inspectors reviewed a sample of PBAPS's submittals for the four Barrier Integrity PIslisted 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. Findings
No findings of significance were identified..2Emergency Preparedness (EP) PIs (71151 - 3 Samples)
a. Inspection Scope
The 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. Findings
No findings of significance were identified.4OA2Identification and Resolution of Problems (71152 - 1 Sample)
.1 Routine Review of Items Entered Into the CAP
a. Inspection Scope
As 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 PBAPS's CAP. This review was accomplished by selectively reviewing copies of IRs and accessing PBAPS's computerized database.
b. Findings
No findings of significance were identified.
15Enclosure.2Review of Operator Work-Arounds (OWAs) (71152 - 1 Work-Around Sample)
a. Inspection Scope
As 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, OP-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 PBAPS:*Unit 3 Steam Jet-Air Ejector (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:
- (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. Findings
No 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 IRs that Exelon initiated at PBAPS 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 PBAPS'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 PBAPS 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 Scope
URI 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. PBAPS'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. Findings
Introduction.
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. PBAPS's 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 RCR identified that a root cause for the configuration error was that standards,policies, and administrative controls (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 PBAPS'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. Findings
Introduction.
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 PBAPS 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.The 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 <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> 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 EDGs 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 PBAPS 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 EDGs.
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 PBAPS's CAP as IR 604364, this finding is being treated as an 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 Scope
The inspectors reviewed corrective action documents listed in the Attachment to thisreport, and discussed the events surrounding the failure of the DDFP 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."
b. Findings
Introduction.
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 DDFP, 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 DDFP 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 DDFP 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 DDFP 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 DDFP 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 DDFP cooling water isolation valves on May 23, 2007, resulting in the DDFP 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 PBAPS's CAP as IR 633532, this violation is being treated as an NCV, consistent with Section VI.Aof 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 SRVs. 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.
22Enclosure.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 HPSW 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 HPSW 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 132' 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 PBAPS 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.
.2 Annual 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 PBAPS 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 ML071000066. 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 NCVs.*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 18 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 PBAPS's 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 PBAPS's 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 PBAPS's CAP as IR 539277. As documented in Section 4OA3.4, 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 SRVs.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
- 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 Personnel
- F. Bower, Senior Resident Inspector
- M. 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 DISCUSSED
Opened None.Opened and
Closed
- 05000278/FIN-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 4OA3.1)05000277, 278/2007003-02NCVMissed Procedure Step Resulted inUnplanned Overloading of the E-3
- EDG (Section 4OA3.2)
- A-2Attachment05000277, 278/2007003-03NCVInadequate Procedure AdherenceResults in Damage to the DDFP
(Section 4OA3.3)
Closed
- 05000278/FIN-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 4OA3.1)05000277, 278/2007003-02NCVMissed Procedure Step Resulted inUnplanned Overloading of the E-3
- EDG (Section 4OA3.2)
- A-2Attachment05000277, 278/2007003-03NCVInadequate Procedure AdherenceResults in Damage to the DDFP
(Section 4OA3.3)
Discussed
None.
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse WeatherWC-AA-107, Revision 4, Seasonal ReadinessOP-AA-108-111-1001, Revision 2, Severe Weather and Natural Disaster Guidelines
- OP-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 Operations
Section 1R04: Equipment AlignmentCOL 52A.1.A-3, Revision 12,
- E-3 Diesel Generator Normal StandbySO 53.7.A - App 1, Revision 0, Removal of 220-08 Line from Service
- 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 System
Section 1R05: Fire ProtectionPF-63, Revision 1, Prefire Strategy Plan Unit 3 Reactor Bldg.
- RCIC Room, 88' ElevationPF-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' ElevationPF-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', Fire Zone 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 164
==Section 1R12: Maintenance EffectivenessIR
- 607398, Functional Failure of 3AE015 During '4T4' Breaker FireIR
- 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, ITE Breaker Found With No Rejection Tab
- IR 634971, ITE Breaker Found With No Rejection Tab
- IR 634962, ITE Breaker Found With No Rejection Tab
- IR 634964, ITE Breaker Found With No Rejection Tab
- IR 634966, ITE Breaker Found With No Rejection Tab
- IR 634965, ITE Breaker Found With No Rejection Tab
- IR 600797, 2007 Buried Pipe Program Inspections
- A-4AttachmentIR
- 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
- 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 TemplateIR
- 620785, Continuous Venting of the Recirculation Seals not EvaluatedAO 2A.16-3, Revision 2, Manual Adjustment of Recirculation Pump Seal Second Stage PressureSO 2A.1.C-3, Revision 10, Operation of the Recirculation Pump Seal Purge System
- A1439223, 3AP034: Seal Hi Temp Alarm & Hi 2
nd Stage PressureACMP - Unit 3, 3 'B' Recirculation Pump Seal Increasing Second Stage Seal Pressure
- A1613202, 3 'B' Recirculation Pump 2
nd Stage Seal PressureIR
- 619609, 3 'B' Recirculation Pump 2
nd 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
- OP-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 2SUE 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, 3CK001 - 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 OperationIR
- 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 System
Section 1R15: Operability EvaluationsIR 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 PressureOTDM No. 07-02, 3 'A' RRP Seal Temperatures - Re-align CRD Suction Source fromCondensate System to U3 CST, dated 04/20/2007AR 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-2-02-117DH Sensing Line Leakage, datedMay 24, 2007A1615458, Small Leak on
- DPT-2-02-117D Line Snubber Threaded Cap
- C0221439, Replace Snubber During an Outage
- PB
Section 1R19: Post-Maintenance TestingAR A1610537, High Lube Oil Temperature Alarm During
- E-2 EDG RunR1049367, Unit 3
- 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, 2BG002, Replace EndbellMA-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, 3CP343: 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 Off
Section 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, 1997
==Section 1R23: Temporary Plant ModificationsIR
- 618478, Provide Supplemental Cooling to the 3 'A'==
- RR Seal Purge LineIR
- 625092, Equipment Discovered on Floor Hatch H11 in Unit 3 Reactor Building
- C0221034, TCCP 07-00172, Install Cooling Unit
- AR A1613094, Provide Supplemental Cooling to the 3 'A' RR Seal Purge Line
- SO.005-3, Revision 1, Routine Inspection of the 3 'A' Recirculation Seal Purge SupplementalCooling System
Section 1EP2: 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
- EP-MA-121-1004 "Exelon East ANS Corrective Maintenance," Revision 4
- EP-MA-121-1005 "Exelon East ANS Preventive Maintenance Program," Revision 3
- EP-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 SirensIRs:
- 00433494 00565056
- 00352078
- 00597065
- 00481763 00451040
- 00520830
- 00521321
- 00533157 00541478
- 00596641
- A-7Attachment
Section 1EP4: 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"
- EP-AA-1000, Revision 17
- 06-33 "EP Plan Administration"
- EP-AA-120, Revision 7
- 06-96 "Emergency Preparedness Advisory Committee"
- EP-AA-120-1004, Revision 0
- 06-97 "Quarterly Satellite Phone Test"
- EP-MA-124-1004, Revision 0
- 06-99 "EP Fundamentals"
- EP-AA-1101, Revision 3
- 06-101"Exelon East ANS Program"
- EP-MA-121-1002, Revision 4
- 06-102"Exelon East ANS Corrective Maintenance Program"
- EP-MA-121-1004, Revision 4
- 06-103"Exelon East ANS Preventative Maintenance Program"
- EP-MA-121-1005, Revision 3
- 06-108"ERO Fundamentals"
- EP-AA-1102, Revision 2
- 06-110"Mid-Atlantic ERO Notification or Augmentation" EP-AA-112-100-F-07
- EP-MA-121-1006, Revision 407-12"ANS Siren Monthly Test"
- RT-E-101-901-2, Revision 8
- 07-18"Radiological Emergency Plan Annex for PBAPS"
- EP-AA-1007, Revision 14
- EP-MA-121-1006, Revision 5
Section 1EP5: Correction of Emergency Preparedness WeaknessesEP-AA-125 "Emergency Preparedness Self Evaluation Process," Revision4LS-AA-126 "Self-Assessment Program," Revision 5
- LS-AA-126-1001 "Focused Area Self-Assessments," Revision 4
- Unusual Event Evaluation Reports dated 10/4/06, 11/21/06, 4/16/07
- ASSAs:
- 547869, 565747-04
- NOSA:-PEA-06-03 dated 4/13/06
- NOSA-PEA-07-04 dated 5/9/07
- A-8AttachmentIRs:
- 00433494
- 00565056
- 00352078
- 00481763
- 00451040 00520830
- 00533157
- 00541478
- 00596641 00597065
- 00521321Section 1EP6: Drill EvaluationPeach Bottom Atomic Power Station, May 15, 2007, Off-Year Exercise, Drill ScenarioPeach Bottom Atomic Power Station May 15
th, 2007 Off-Year Exercise Report datedJune 14, 2007IR
- 630584, Enhancement Opportunity from May 2007 EP Drill
- IR 629910, Late State\Local Notification Made During an EP Drill
- IR 629970, EAL Classification During Drill Not Timely Quick Human Performance Investigation Report, PB EAL Classification During Drill Not Timely, 05/15/07Quick Human Performance Investigation, Repetitive Issue With Not Completing State/Local Notifications on Time, 5/15/07 Section 2PS2 : Radioactive Material Processing and Transportation10 CFR Part 61 Sampling and Analysis Results (Waste Streams)Radioactive Material Shipping Documentation Radioactive Shipping Container Certifications Audit Template: Chemistry, Radwaste, Effluent and Environmental Monitoring Handling, Storage and Shipping Topical Report, Mobile In-container De-watering and Solidification System
- DOT-Type A, Test and Evaluation for Type A Packaging Waste Disposal Facility State Licenses Training Program - DOT/79-19 Training for Support of Radioactive and Asbestos Shipments Training Program - Site Specific Portion of Radioactive Material Shipping Training Program Training Program - Shipper Refresher Type B Cask Handling and Loading Procedures
- RT-W-020-980-2, Updating Radwaste Classification Computer Programs
- RP-AA-605, 10
- CFR 61 Compliance Program
- RP-AA-605, 10
- CFR 61 Program
- RP-PB-605-1001, Peach Bottom 10
- CFR 61 Sampling
Section 4OA1: Performance Indicator VerificationLS-AA-2001, Revision 6, Collecting and Reporting of
- NRC Performance Indicators DataLS-AA-2090, Revision 4, Monthly Data Elements for NRC RCS Specific Activity
- LS-AA-2100, Revision 5, Monthly Data Elements for NRC RCS Leakage
- ST-O-020-560-2, Reactor Coolant Leakage Test (sample of completed test records)
- ST-O-020-560-3, Reactor Coolant Leakage Test (sample of completed test records)
- ST-C-095-864-2, Off Gas Monitor Response and Release Rate Verification by a Grab Sample
- A-9AttachmentST-C-095-864-3, Off Gas Monitor Response and Release Rate Verification by a Grab SampleST-C-095-820-2, Determination of Dose Equivalent µCi/g I-131 in Primary Coolant
- ST-C-095-820-3, Determination of Dose Equivalent µCi/g I-131 in Primary Coolant
- CH-407, Sampling of Reactor Water
- Public Notification System PI data, April 2006 - March 2007
Section 4OA2: Problem Identification and Resolution577381, Operator Failed to Perform Procedure Step581258, Page 12 of
- ST-O-098-01N-2 Discovered Misplaced
- 568038, SBLC System Inoperable Resulting from Dedicated EO Leaving Area
- 565945, 4 kV Undervoltage Relay Failure and No IR's Written
- 569879, 4 kV Undervoltage Relay Failure and No IR's Written
- 576826, NOS Rated PB OPS Yellow For 4Q06
- 581376, Test Aborted: "ECT Portable Pump Operability" RT-O-48B-275-2
- 584506, Through Wall Leak Found on ESW Piping585680, Unit 3 'D' RHR Exceeded the Original Dose Estimate587171,
- CHK-0-33-515A Not Seated Causes
- ST-0-033-300-2 To Be Aborted
- 588335, Timeliness/Response to ESW Piping Issue
- IR 584506
- 588800, Weld Verification Deficiency
- 590373, Trng: FME Training Unexcused Absence
- 590573, E/S 3-17-477 Power Supply Failed Following Swap of 3 'B' RPS
- 596641, Unusual Event Notification to York County Was > 15 Minutes
- 2264, Mid-Cycle Performance Gap - Self Assessment
- 606458, Training: PIMS Code Improperly Granted
- 607064, Temperature Recorder
- TR-0558 not Functional (Discharge Canal)
- 615413, Non-Safety-Related Piece Part Installed in Diesel Generator
- 23697, Scaffold Taken to Complete in PIMS But Was Not Removed
- 29910, Late State/Local Notification Made During an EP Drill
- 29970, EAL Classification During Drill Not Timely
- 26534, Equipment Not Protected As Required
- 596616, Fault at 3 'B' Iso-Phase Cooler Fan Breaker in '4T4' Load Center
- 633532, DDFP/ Engine Trip
- 604364, Human Error Results in E-3 EDG Overload and E-33 Breaker Trip
Section 4OA3: Event FollowupSpecial Event Procedure (SE)-4, Flood, Revision 21
- IR 563253, External Flood Vulnerability - Circulating Water Pump Structure
- IR 554800, External Flood Vulnerability Found for EDG Building
- IR 520322, E-3 EDG Fire at Roof Exhaust Penetration
- A-10AttachmentST-O-37D-370-2, Revision 25, DDFP Operability TestST-O-37D-340-2, Revision 10, DDFP Flow Rate Test
- ST-O-37D-340-2, Revision 12, DDFP Flow Rate Test
- ST-M-37D-380-2, Revision 3, DDFP Inspection
- NOM-C-7.1, Revision 2, Procedure Use
- 280-E-8, Revision 16, Single line Meter and relay Diagram, Standby Diesel Generators and4160 Volt Emergency Power System, Unit 26280-E-1615, Revision 64, Single Line Meter and relay Diagram, E-124 and E-224 Emergency Load Centers, E-124-R-C and E-224-R-C Reactor Motor Control Centers, and
- E-124-T-B and E-224-T-B Turbine Motor Control Centers, 440 Volt, Unit 2Peach Bottom Atomic Power Station Fire Protection Plan, Revision 15
- A-11AttachmentIssue ReportsIR 00633037IR 00633453
- IR 00634313IR 00634585IR 00634709
- AR 00635028AR 00635257AR 00635267
LIST OF ACRONYMS
ADAMSAgency-wide Documents Access and Management SystemADSautomatic depressurization system
ANSAlert and Notification System
ARaction request
BTPsbranch technical positions
CAPcorrective action program
CFRCode of Federal Regulations
CRcondition report
CRDcontrol rod drive
CSTcondensate storage tank
DBDsDesign Basis Documents
DDFPdiesel-driven fire pump
DEPDrill and Exercise Performance
DOTDepartment of Transportation
DRPDivision of Reactor Projects
EALemergency action level
EDGemergency diesel generator
EPemergency preparedness
EROemergency response organization
ESWemergency service water
HPCIhigh pressure coolant injection
HPSWhigh pressure service water
HCUhydraulic control unit
IMCInspection Manual Chapter
IPInspection Procedure
IRissue report
kVkilovolt
LERslicensee event reports
MRMaintenance Rule
MSmitigating system
NCVnoncited violation
NEINuclear Energy Institute
NRCNuclear Regulatory Commission
OCsoperator challenges
OWAsoperator work-arounds
PARprotective action recommendation
PARSPublicly Available Records
PBAPSPeach Bottom Atomic Power Station
PCIVprimary containment isolation valve
PCPProcess Control Program
2AttachmentPIperformance indicatorPMTpost-maintenance testing
PROplant reactor operator
RBreactor building
RCICreactor core isolation cooling
RCRroot cause report
RFPreactor feed pump
RGRegulatory Guide
RHRresidual heat removal
RRPreactor recirculation pump
RTPrated thermal power
SDPsignificance determination process
SJAEsteam jet-air ejector
SPACstandards, policies, and administrative controls
SOsystem operating
SSCsstructures, systems, and components
SRVsafety relief valve
STssurveillance tests
TRMTechnical Requirements Manual
TRTtroubleshooting, rework and testing
TSTechnical Specification
TSCtechnical support center
UEunusual event
URIunresolved item
UFSARUpdated Final Safety Analysis Report
USQunreviewed safety question