IR 05000277/2007003
Download: ML072120599
Text
July 30, 2007
Mr. Christopher M. CranePresident and CNOExelon NuclearExelon Generation Company, LLC200 Exelon Way Kennett Square, PA 19348
SUBJECT: PEACH BOTTOM ATOMIC POWER STATION - NRC INTEGRATEDINSPECTION 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 interviewedpersonnel. 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, ChiefReactor 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/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/Paul G. Krohn, ChiefReactor Projects Branch 4 Division of Reactor ProjectsDocket Nos.:50-277, 50-278License Nos.:DPR-44, DPR-56
Enclosures:
Inspection Report 05000277/2007003 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
.....................................................iii
REPORT DETAILS
...........................................................1REACTOR SAFETY..........................................................11R01Adverse Weather Protection........................................11R04Equipment Alignment..............................................2
1R05 Fire Protection...................................................3
1R11 Licensed Operator Requalification Program.............................4
1R12 Maintenance Effectiveness.........................................41R13Maintenance Risk Assessments and Emergent Work Control...............51R15Operability Evaluations............................................51R19Post-Maintenance Testing..........................................61R22Surveillance Testing...............................................6
1R23 Temporary Plant Modifications.......................................7
1EP2Alert and Notification System (ANS) Evaluation..........................71EP3Emergency Response Organization (ERO) Staffing and Augmentation System.8 1EP4Emergency Action Level (EAL) and Emergency Plan Changes..............81EP5Correction of Emergency Preparedness Weaknesses.....................91EP6Drill Evaluation...................................................9RADIATION SAFETY........................................................102PS2Radioactive Material Processing and Transportation.....................10OTHER ACTIVITIES.........................................................134OA1Performance Indicator (PI) Verification...............................134OA2Identification and Resolution of Problems.............................144OA3Event Followup.................................................16 4OA5Other Activities..................................................23 4OA6Meetings, Including Exit...........................................23 4OA7Licensee-Identified Violations......................................23
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
.................................................A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
............................A-1
LIST OF DOCUMENTS REVIEWED
...........................................A-2
LIST OF ACRONYMS
......................................................A-10
- OF [[]]
- FINDIN [[]]
GSIR 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's program for overseeing the safe
operation of commercial nuclear power reactors is described in
- NUR [[]]
EG-1649, "Reactor Oversight
Process," Revision 4, dated December 2006.
- A.NRC -Identified and Self-Revealing FindingsCornerstone: 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
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
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
- PBA [[]]
PS personnel did not follow
procedures when the E-3 EDG was placed in the isochronous load control mode
with the E-3
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
- DD [[]]
cooling water that subsequently damaged the engine during operations without
cooling water. The failure to use a procedure for cleaning and restoring the
- DD [[]]
FP cooling waterstrainer was a more than minor finding because it was associated with the
degradation of a fire protection feature, in that, the
- DD [[]]
FP was rendered inoperable
by damage to the engine. 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
- DD [[]]
FP 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
- DD [[]]
FP cooling water
strainer (IMC 0305 aspect
OA3.3)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.
EnclosureREPORT
- DETAIL [[]]
SSummary 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
'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.
- 1.REACT [[]]
- OR [[]]
- SAFET [[]]
YCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 System Sample; 1 Site Sample).1Summer Seasonal Readiness a.Inspection ScopeThe inspectors performed one seasonal readiness sample that included a review of threeventilation systems. Specifically, the inspectors reviewed the procedures listed in
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.FindingsNo findings of significance were identified.
2Enclosure.2Adverse Weather Event Review a.Inspection ScopeOn June 13, 2007, a tornado warning was issued for an adjacent county. The inspectorsreviewed
- PBA [[]]
PS'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
- PBA [[]]
PS's personnel
consulted procedure
PB-108-111-1001, "Preparation for Severe Weather," increased
the online risk assessment to "Yellow," and subsequently implemented procedure
AO 53.2-0, "Equipment Checks After a Thunderstorm." b.FindingsNo findings of significance were identified.1R04Equipment Alignment (71111.04Q - 3 Partial Walkdown Samples).1Partial Walkdown a.Inspection ScopeThe inspectors performed a partial walkdown of three systems to verify the operability ofredundant or diverse trains and components when safety-related equipment was
inoperable. The inspectors performed walkdowns to identify any discrepancies that could
impact the function of the system and potentially increase risk. The inspectors reviewed
applicable operating procedures, walked down system components, and verified that
selected breakers, valves, and support equipment were in the correct position to support
system operation. The inspectors also verified that
- PBA [[]]
PS had properly identified and
resolved equipment alignment problems that could cause initiating events or impact the
capability of mitigating systems or barriers and entered them into the CAP. The three
systems reviewed were: *E-3 Diesel Generator and 3 Startup Transformer with the 2 Startup TransformerOut-of-Service;*Unit 2 Reactor Core Isolation Cooling (RCIC) with Unit 2 High Pressure CoolantInjection (HPCI) Out-of-Service; and*'B' Emergency Service Water (ESW) with 'A'
- ESW Out-of-Service. b.FindingsNo findings of significance were identified..2Complete System Walkdown (71111.04S - 1 Sample) a.Inspection ScopeDuring the week of June 25, 2007, the inspectors performed one complete Unit 2 highpressure service water (
HPSW) system walkdown of the accessible portions of the
3Enclosuresystem. The full walkdown was performed to identify any discrepancies which couldimpact the Unit
- 2 HP [[]]
SW system function. The inspectors reviewed system operating
procedures, piping and instrumentation drawings, walked down control system
components, and verified that circuit breakers and valves were in the appropriate
positions. b.FindingsNo findings of significance were identified.1R05Fire Protection (71111.05Q - 10 Samples)Fire Protection - Tours a.Inspection ScopeThe inspectors reviewed
TRM), and the respective pre-fire action plan procedures to determine the required fire
protection design features, fire area boundaries, and combustible loading requirements for
the areas examined during this inspection. The fire risk analysis was reviewed to gain risk
insights regarding the areas selected for inspection. The inspectors performed
walkdowns of ten areas to assess the material condition of active and passive fire
protection systems and features. The inspection was also performed to verify the
adequacy of the control of transient combustible material and ignition sources, the
condition of manual firefighting equipment, fire barriers, and the status of any related
compensatory measures. The following ten fire areas were reviewed for impaired fire
protection features:*Unit 3 Reactor Building (RB),
- RC [[]]
IC Room, 88' Elevation (Fire Zone 63);*Standby Gas Treatment Room, Radwaste Building, 91'6" Elevation (Fire Zone 70);
- Unit
CRD) Equipment and West Corridor (Fire Zone 13H);*Unit 3 Refuel Floor (Fire Zone 55);
- Unit 3 'A' and 'C' Core Spray Rooms (Fire Zones 13D & 13E);
- Unit 2 Emergency Battery/Switchgear Rooms (Fire Zone 127);
- Unit
- 2 RC [[]]
IC (Fire Zone 60);
- Unit 2 Main Transformer Yard (Fire Zone 151);
- 2 Startup Switchgear Building (Fire Zone 164); and
- Diesel Generator Building, 127' Elevation (Fire Zone 132). b.FindingsNo findings of significance were identified.
4Enclosure1R11Licensed Operator Requalification Program (71111.11Q - 1 Sample) Resident Inspector Quarterly Review a.Inspection ScopeOn June 12, 2006, the inspectors observed operators in
- PBA [[]]
PS's simulator duringlicensed operator requalification training to verify that operator performance was adequate
and that evaluators were identifying and documenting crew performance issues. The
inspectors verified that performance issues were discussed in the crew's post-scenario
critiques. The inspectors also observed operator implementation of procedures. The
inspectors discussed the training, simulator scenarios, and critiques with the operators,
shift supervision, and the training instructors. The evaluated scenario observed for this
one sample involved the events listed below: *Small Break Loss of Coolant Accident; and*An Anticipated Transient Without Scram. b.FindingsNo findings of significance were identified.1R12Maintenance Effectiveness (71111.12Q - 2 Samples) a.Inspection ScopeThe inspectors reviewed two samples of
- PBAPS 's evaluation of degraded conditionsinvolving safety-related structures, systems, and components (
SSCs) for maintenance
effectiveness during this inspection period. The inspectors reviewed
- PBA [[]]
PS's
implementation of the Maintenance Rule (MR), and verified that the conditions associated
with the referenced condition reports (CRs) were evaluated against applicable MR
functional failure criteria as found in the licensee's scoping documents and procedures.
The inspectors also discussed these issues with system engineers and MR coordinators
to verify that they were tracked against performance criteria and that the systems were
classified in accordance with MR implementation guidance. Documents reviewed during
the inspection are listed in the Attachment. The following conditions were reviewed:*Issue Report (IR) 587171,
- ESW [[]]
ST-O-033-300-2 to be Aborted; and*IR 622560, Maintenance Preventable Functional Failure for Loss of '4T4' 480 VoltLoad Center. b.FindingsNo findings of significance were identified.
5Enclosure1R13Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 Samples) a.Inspection ScopeThe inspectors evaluated
AA-101, "On-line Work Control Process," was also reviewed. This inspection
included reviews of
- PBA [[]]
PS's use of the Paragon online risk monitoring software. The
inspectors reviewed equipment tracking documentation, daily work schedules, and
performed plant tours. The following activities selected were based on plant maintenance
schedules and systems that contributed to risk. The inspectors completed eight
evaluations of maintenance activities on the following:*Troubleshooting, Rework and Testing (TRT) Control Form No. 07-18, Monitor3 'A' Recirculation Pump Seal Parameters During Recirculation Pump Speed
Changes;*TRT No.07-020, Re-align
CST) from the Condensate System;*WO C0220911, Calibrate, Repair & Replace E-2 EDG Temperature Switch;
- WO A1613202, 3 'B' Recirculation Pump 2nd Stage Seal Pressure;*IR 623723, Bolt and Heli-coil Found Damaged at Disassembly on 00T634;
- IR 626534, Equipment Not Protected As Required;
- WO R0736769-01, Core Spray Loop 'A' Full Flow Test Valve Operator,MO-2-14-026A-OP, Perform Motor Operator Preventive Maintenance; and*IR 542109, 2 'C' Service Air Compressor Trip.Additionally, the inspectors verified that an inspector-identified issue,
PBAPS's CAP. b.FindingsNo findings of significance were identified.1R15Operability Evaluations (71111.15 - 5 Samples) a.Inspection ScopeThe inspectors reviewed five issues to assess the technical adequacy of the evaluations,the use and control of compensatory measures, and compliance with the licensing and
design bases. Associated adverse condition monitoring plans, engineering technical
evaluations, and operational and technical decision making documents were also
reviewed. The inspectors verified these processes were performed in accordance with
the applicable procedures. The inspectors used
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
IR 615413);*Rising 3 'A' Reactor Recirculation Pump (RRP) #2 Seal Temperature (IR 617988);
- Provide Supplemental Cooling to the 3 'A'
IR 618478);
6EnclosureTarget Rock Safety/Relief Valve (SRV) Seal Welds: Potential Code Issue(IR 628251); andSmall Leak on 2 'B' Main Steam Line Differential Pressure Instrument LineSnubber Threaded Cap (IR 627026). b.FindingsNo findings of significance were identified.1R19Post-Maintenance Testing (71111.19 - 7 Samples) a.Inspection ScopeThe inspectors observed selected portions of post-maintenance testing (PMT) activitiesand reviewed completed test records. The inspectors observed whether the tests were
performed in accordance with the approved procedures and assessed the adequacy of
the test methodology based on the scope of maintenance work performed. In addition,
the inspectors assessed the test acceptance criteria to evaluate whether the test
demonstrated that the tested components satisfied the applicable design and licensing
bases and the TS requirements. The inspectors reviewed the recorded test data to verify
that the acceptance criteria were satisfied. The inspectors reviewed seven PMTs
performed in conjunction with the following maintenance activities:*WO C0220911, Calibrate, Repair & Replace E-2
WO R1049367, Unit 3 Hydraulic Control Unit (HCU) 50-43: HCU Overhaul;
- WO R1017055,
DR) Diesel Engine 6YR Overhaul;
- WO C0216504,
PS-2-13-067-01), Replace Pressure Switch;
WO C0221445, Inspect/Repair/Replace Unit 2 'C' Main Steam Line Radiation
Monitor (RIS-2-17-251C);*WO C0215740, Replace Unit 2 'B' Reactor Protection System Motor Generator Set Endbell; and*WO R0629147, Perform Motor Control Unit Inspection on the 'C' Glycol Pump. b.FindingsNo findings of significance were identified.1R22Surveillance Testing (71111.22 - 5 Samples) [3 Routine Samples;
RCS) Leakage Sample] a.Inspection ScopeThe inspectors reviewed and observed portions of selected surveillance tests (STs), andcompared test data with established acceptance criteria to verify the systems
demonstrated the capability of performing the intended safety functions. The inspectors
also verified that the systems and components maintained operational readiness, met
applicable TS requirements, and were capable of performing the design basis functions.
The five STs reviewed and observed included:
7Enclosure*ST-O-023-301-3,
- HPCI [[Pump, Valve, Flow and Unit Cooler Functional andIn-Service Test []]
IST Sample];*ST-O-020-560-2 & 3, Reactor Coolant Leakage Test [RCS Leakage Sample];
- ST-I-60A-230-3, Linear Power Range Monitor Gain Calibration;
- SI2T-MIS-8547-C1CQ, Calibration/Functional Check of Channel 'C' Group 1, 4and 5 of Primary Containment Isolation Valve (PCIV) Logic for
ST-R-003-485-3, CRD Scram Insertion Timing of Selected Control Rods.b.FindingsNo findings of significance were identified.1R23Temporary Plant Modifications (71111.23 -1 Sample) a.Inspection ScopeThe inspectors reviewed one temporary modification to verify that implementation of themodification did not place the plant in an unsafe condition. The review was also
conducted to verify that the design bases, licensing bases, and performance capability ofrisk significant SSCs had not been degraded as a result of the modification. The
inspectors verified the modified equipment alignment through control room
instrumentation observations; the
WO reviews; and plant
walkdowns of accessible equipment. The following temporary modification was reviewed:*TCCP 07-00172, Install Cooling Unit to Assist 3 'A'
- RRP Seal Cooling. b.FindingsNo findings of significance were identified.Cornerstone: Emergency Preparedness1
- EP 2Alert and Notification System (ANS) Evaluation (71114.02 - 1 Sample) a.Inspection Scope An onsite review was conducted to assess the maintenance and testing of the
ANS. During this inspection, the inspectors interviewed emergency preparedness (EP)
staff responsible for implementation of the
IRs pertaining
to the ANS were reviewed for causes, trends, and corrective actions. The inspectors
further discussed with
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
IP)
71114, Attachment 2. Planning standard, 10 CFR 50.47(b)(5) and the related
requirements of 10 CFR 50, Appendix E were used as reference criteria.
8Enclosure b.FindingsNo findings of significance were identified.1EP3Emergency Response Organization (ERO) Staffing and Augmentation System (71114.03 - 1 Sample)
a.Inspection Scope A review of Peach Bottom's
ERO 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
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
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 [[]]
CFR 50, Appendix E were used as
reference criteria. b.FindingsNo findings of significance were identified.
PBAPS's determination, in accordance with
CFR 50.54(q), that the changes resulted in no decrease in effectiveness of the Plan,
and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and
Appendix E to 10 CFR 50. The inspectors conducted a sampling review of the
Emergency Plan changes, and changes to the lower-tier Emergency Plan implementing
procedures, to evaluate the changes for potential decreases in effectiveness of the
Emergency Plan. However, this review was not documented in a safety evaluation report
and does not constitute formal NRC approval of the changes. Therefore, these changes
remain subject to future NRC inspection in their entirety. b.FindingsNo findings of significance were identified.
9Enclosure1EP5Correction of Emergency Preparedness Weaknesses (71114.05 - 1 Sample) a.Inspection Scope The inspectors reviewed a sampling of self-assessment procedures and reports to assessPBAPS's ability to evaluate their performance and programs. The inspectors reviewed a
sampling of IRs from July 2006 through May 2007, initiated by Exelon Nuclear at Peach
Bottom from drills, self-assessments, and audits. Other drill reports reviewed included:
medical/health physics, fire, integrated, and call-in. Additionally, the inspectors reviewed
the three UE Evaluation Reports generated since the last inspection, and audits for 2006
and 2007 required by 50.54(t). This inspection was conducted in accordance with
- NRC [[]]
71114, Attachment 5. Planning standard, 10 CFR 50.47(b)(14) and the related
requirements of
- 10 CFR 50, Appendix E were used as reference criteria. b.FindingsNo findings of significance were identified.1
EP6Drill Evaluation (71114.06 - 1 Sample)Off-Year Exercise (Drill) a.Inspection ScopeThe inspectors conducted this inspection to assess: training quality and conduct;emergency plan procedure implementation; facility and equipment readiness; personnel
performance in drills and exercises; organizational and management changes; and
communications equipment readiness. The primary focus of this inspection was to verify
- PBA [[]]
PS's critique of classification, notification, and protective action recommendation
(PAR) development.On May 15, 2007, the inspectors observed a full scale drill. The primary focus of thisinspection was to verify
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
TSC. The
inspectors observed
PBAPS's identification of
weaknesses and deficiencies. The inspectors compared
- PBA [[]]
PS's identified issues
against the inspectors' observations to determine whether
- PBA [[]]
PS adequately identified
problems and entered them into the CAP. This inspection activity represented one
10Enclosuresample. The documents and procedures reviewed during the inspection are listed in theAttachment. b.FindingsNo findings of significance were identified.2.RADIATION
- PS 2Radioactive Material Processing and Transportation (71122.02 - 5 Samples).1Inspection Planning/In-Office Inspection a.Inspection ScopeThe inspectors reviewed the solid waste system description in the
- UFS [[]]
AR and recentradiological effluent release reports for information on the types and amounts of
radioactive waste. The inspectors reviewed Exelon's audit program in the area of radioactive wastecharacterization, transportation, and disposal. The inspectors also reviewed the status of
the
PS2.6) b.FindingsNo findings of significance were identified. .2Radioactive Waste System Walkdown a.Inspection ScopeThe inspectors walked down accessible portions of the station's radioactive liquid andsolid waste collection, processing, and storage systems and locations to determine if:
systems and facilities were consistent with descriptions provided in the
- UFS [[]]
AR; to
evaluate their general material conditions; and to identify changes made to systems.
Areas visually inspected included tank and pump rooms, the de-watering facility, in-plant
and outside waste storage areas, outside tank areas, and the low level-waste storage
facility. Visual inspection records and previous surveys were also reviewed. The
inspector also discussed operation of the radwaste systems with cognizant licensee
personnel.The inspectors reviewed the status of any non-operational or abandoned radioactivewaste process equipment; the adequacy of administrative and physical controls for those
systems; changes made to radioactive waste processing systems and potential
radiological impact, including conduct of safety evaluations of the changes, as necessary.
11EnclosureThe inspectors reviewed the current processes for transferring radioactive waste resin andsludge to shipping containers and mixing and sampling of the waste, as appropriate, to
evaluate waste mixing, adequacy of sampling, and the methodology for waste
concentration averaging. The inspector also reviewed radioactive waste and material
storage and handling practices; sources of radioactive waste at the station (waste
streams) and processing (as appropriate) and handling of the waste; and the general
condition of facilities and equipment. The review was against criteria contained in the station's
PCP), and applicable station procedures. b.FindingsNo findings of significance were identified..3Waste Characterization and Classification a.Inspection ScopeThe inspector reviewed the following matters:
- Radio-chemical sample analysis results for radioactive waste streams;*Development of scaling factors for difficult to detect and measure radionuclides;
- Methods and practices to detect changes in waste streams;
- Classification and characterization of waste relative to
BTPs) on wasteclassification, concentration averaging, waste stream determination, and sampling
frequency;*Current waste streams and their processing relative to descriptions contained inthe
- PCP ; *Current processes for transferring radioactive waste resin and sludge dischargesinto shipping/disposal containers to determine adequacy of sampling; *Revisions of the
UFSAR to reflect changes (as appropriate); and
- Waste processing topical report (de-watering).The inspector discussed the adequacy of samples collected from the waste transfer andde-watering system.The review was against criteria contained in
NRC BTPs, and Exelon procedures. b.FindingsNo findings of significance were identified.
2Enclosure.4Shipment Preparation a.Inspection ScopeThe inspector observed a non-exempt radioactive material shipment (PM-07-057) inpreparation. The inspector reviewed associated transportation documents, reviewed
radiological surveys to support transportation, reviewed license requirements, and
discussed preparation with cognizant Exelon personnel. The inspector also reviewed
personnel training relative to
CFR 172, Subpart H. The
inspector reviewed and discussed technical training presented to workers. The inspector
verified that a training program was provided to personnel responsible for the conduct of
radioactive waste processing and radioactive waste shipping activities. b.FindingsNo findings of significance were identified. .5Shipment Records and Documentation a.Inspection ScopeThe inspector selected and reviewed the records associated with six non-exceptedshipments of radioactive material made since the previous inspection in this area
(Shipment Nos.
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
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
DOT requirements, as contained in 49 CFR 170-189 for the above areas;
station procedures; applicable disposal facility licenses; and applicable Certificates ofCompliance or vendor procedures for various shipping casks.The inspector also selectively reviewed the 2006 Annual Radioactive Effluent ReleaseReport, relative to types and quantities of radioactive waste shipped offsite and relative to
changes to the
- PCP. [[b.FindingsNo findings of significance were identified..6 Audits and Assessments of Radioactive Waste Handling a.Inspection ScopeThe inspector reviewed audits and assessments of the radioactive waste handling,processing, storage, and shipping programs, including the]]
PCP. The inspector also
reviewed selected corrective action documents written since the previous inspection. The
following documents were reviewed:*Chemistry, Radwaste, and Process Control Audit, (NOSA-PEA-06-04 (IR 476157),May 3, 2006; *Self-Assessment,
- AS [[]]
SA-565928 A05, May 14, 2007; and
- Issue Reports (IRs) 632879, 626897, 626873, 618653, 612012, 605803,592478486694, 240959, 642483, 642097, 642491, 632526, 486694. The review was against criteria contained in
- CFR 71.101, andapplicable station audit and surveillance procedures. b.FindingsNo findings of significance were identified.4.OTHER
- ACTIVI [[]]
PI) Verification (71151 - 7 Samples)
.1Barrier Integrity
PI definitions and the
guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment
Indicator Guideline," Revision 4, and Exelon procedure
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
reviewed were:*Unit 2 and Unit
- 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
NEI 99-02, Revision 4, "Regulatory Assessment
Performance Indicator Guidelines." b.FindingsNo findings of significance were identified.4OA2Identification and Resolution of Problems (71152 - 1 Sample) .1Routine Review of Items Entered Into the
IP 71152, "Identification and Resolution of Problems," and in order to helpidentify repetitive equipment failures, human performance issues or program issues for
follow-up, the inspectors performed routine screening of issues entered into
- PBA [[]]
- PBA [[]]
PS's computerized database. b.FindingsNo findings of significance were identified.
15Enclosure.2Review of Operator Work-Arounds (OWAs) (71152 - 1 Work-Around Sample) a.Inspection ScopeAs required by
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,
AA-102-103, "Operator Work-Around Program." Specifically, the review
was conducted to determine if any OWAs for mitigating systems affected the mitigating
system's safety functions or affected the operator's ability to implement abnormal and
emergency operating procedures. The inspectors reviewed the following open OWAs
being tracked by
SJAE) Suction Valves Fail to Open When Placing theSJAE In-Service (Action Request (AR) A1536806).The inspectors also reviewed the lists of open OCs (deficiencies that are obstacles tonormal plant operations), periodically walked down the panels in the main control room,
and reviewed control room deficiencies to identify and be cognizant of: (1) OWAs that
have not been evaluated by
OWAs that increase the potential for
personnel error, including OWAs that: *Require operations contrary to past training or require more detailed knowledge ofthe system than routinely provided; *Require a change from longstanding operational practices;
- Require operation of a system or component in a manner dissimilar from similarsystems or components;*Create the potential for the compensatory action to be performed on equipment orunder conditions for which it is not appropriate;*Impair access to required indications, increase dependence on oralcommunications, or require actions under adverse environmental conditions; and*Require the use of equipment and interfaces that had not been designed withconsideration of the task being performed.
b.FindingsNo findings of significance were identified..3Semi-Annual Review to Identify Trends (71152 - 1 Semi-annual Trend Sample) .aInspection ScopeAs required by
- IP 71152, Identification and Resolution of Problems, the inspectorsreviewed a list of approximately 5,000
- PBA [[]]
PS from December
1, 2006 through June 1, 2007, to perform the semi-annual PI&R trend review.
Approximately, 30 IRs were reviewed in detail to verify that the issues were adequately
identified, appropriately evaluated and corrected. The inspectors review was focused on
16Enclosurehuman performance issues. The review also included issues documented within
- PBA [[]]
PS'sStation Trend Review for the fourth quarter of 2006 and the first quarter of 2007. b.Assessments and ObservationsAlthough no findings of significance were identified, the inspectors observed that the plantis being challenged by human performance deficiencies. Specifically, procedure
adherence was the aspect of human performance that was most frequently challenged.
Examples are documented in IRs 568038, 577381, 581258, 604364, 596616, 626534 and
633532. Procedure quality was another aspect of human performance that was
challenged. Examples are documented in IRs 635028, 633532, and 600686. However,
the inspectors did not identify any new trends that were not previously identified by
- PBA [[]]
PS under their quarterly Station Trend Review reports. The inspectors noted that the
Station Trend Review report had identified procedure adherence issues as an emerging
trend. The inspectors also noted that improving human performance was identified as
one of five Station Focus areas for 2007.4OA3Event Followup (71153 - 5 Samples) .1(Closed) Unresolved Item (URI)05000277/2007002-04, Incorrect Size Breaker Resultedin a Fire in the '4T4' 480 Volt Load Center a.Inspection ScopeURI 05000277/2007002-04 was opened in
PBAPS had preliminarily determined that the fire resulted from an
apparent mismatch between the ratings of one breaker and its cubicle in the '4T4' 480 volt
load center.
- PBA [[]]
PS's report also documented that operators responded to the equipment
losses caused by the fire by initiating a transient of controlled reactor power reductions to
stabilize the plant at approximately 50 percent of rated power. The URI was opened
pending the
PBAPS's causal
evaluation and corrective actions.
RCR) and the associated
IR 596616 for this event were reviewed to assess the identified issues. The
characterization of this issue as a finding and its risk significance are discussed below.
This
- URI is closed. b.FindingsIntroduction. A Green self-revealing finding was identified for inadequate implementationof
WO instructions to verify the correct breaker frame size during the overhaul of a
compatible spare breaker for installation into the '4T4' 480 volt load center. This condition
resulted in a poor electrical connection between the primary disconnect fingers and the
switchgear bus stabs for one breaker in the '4T4' load center that ultimately resulted in a
fire that led to a plant transient and declaration of an Unusual Event (UE).Description. On February 27, 2007, operators reduced Unit 3 reactor power from 100percent to 50 percent RTP in response to the effects of a fire in the '4T4' 480 volt load
center.
RCR 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 (
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
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)
NRC staffs' characterization of this issue following a
review of
- PBA [[]]
PS's root cause analyses, corrective actions taken or planned, approved
procedures, and other documents. The characterization of this issue as a finding and its
risk significance are discussed below. This
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 [[]]
- 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
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
- PBAPS 's root cause investigation report (IR 604364) tounderstand the underlying causes for this event. The inspectors noted that
- PBA [[]]
PS
identified two root causes for this self-revealing event. First, the plant reactor operator
(PRO) did not adhere to the requirements of
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
EDG to
the isochronous load control mode per the
- ST.T he second root cause for this event was inadequate supervisory oversight during acritical transition between the
ST and SO procedures. Specifically, the peer checker and
the control room supervisor were not directly observing the operation of the E-3 EDG at
the main control room panel during the critical transition between procedures. The
transition between procedures should have been identified as a critical step in the testing
evolution. This breakdown in crew teamwork resulted in the PRO performing a critical
step, without direct oversight, during an infrequently performed test of safety-related
equipment. As a result, no one challenged the
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
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
SDP," Appendix A, "Determining the Significance of Reactor Inspection Findings for
At-Power Situations." The inspectors concluded that this finding affected the Mitigating
Systems Cornerstone and answered "No" to all relevant questions. Specifically, all other
EDG was shorter
than its TS allowed outage time of seven days. Therefore, this finding was considered to
be of very low safety significance (Green).The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance (work practices component) because
- PBA [[]]
PS personnel did not
follow procedures when the E-3 EDG was placed in the isochronous load control mode
with the E-3
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
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 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
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
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
NRC Enforcement Policy: NCV 05000277/2007003-02;05000278/2007003-02, Missed Procedure Step Resulted in Unplanned Overloading
of the E-3
- DDFP a.Inspection ScopeThe inspectors reviewed corrective action documents listed in the Attachment to thisreport, and discussed the events surrounding the failure of the
- DD [[]]
FP with the site fire
protection engineer. The inspectors reviewed Revisions 10 and 12 of ST-O-37D-340-2,
"DDFP Flow Rate Test," and Revision 2 of NOM-C-7.1, "Procedure Use."
20Enclosure b.FindingsIntroduction. A self-revealing Green
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
- DD [[]]
FP, the procedure
directed cleaning of the cooling water strainer, but did not provide specific instructions on
how to perform this task. Without procedure guidance or instructions, operations
personnel performing the
- DD [[]]
FP test closed an upstream hand valve to isolate the strainer
for cleaning. After reassembling the strainer, the operations personnel did not re-open the
hand valve. The cooling water was not properly realigned for service because equipment
manipulations were performed outside of procedure guidance. On May 24, 2007,
ST-O-37D-340-2 was re-performed with the cooling water supply isolated. The engine
was damaged during operation without cooling water as a result of the valve mis-
alignment. The
DDFP flow rate test procedure was revised to include specific instructions
for cleaning the cooling water strainer. The procedure was also revised to include
instructions for monitoring the engine cooling water and lubricating oil parameters during
engine operation. Based on the above, the inspectors determined that manipulating the
- DDFP cooling watervalve without procedure guidance was a performance deficiency. Analysis. The inspectors concluded that the failure to use a procedure for cleaning andrestoring the
DDFP cooling water strainer was a more than minor finding because it was
associated with the degradation of a fire protection feature, in that, the
- DD [[]]
FP was
rendered inoperable by damage to the engine. Traditional enforcement does not apply
since there were no actual safety consequences or potential for impacting the NRC's
regulatory function, and the finding was not the result of any willful violation of
IMC 0609) and determined the finding to be of very low safety significance (Green). The
finding was of low significance due to the motor-driven fire pump remaining operable
during the five days the
- DD [[]]
FP was inoperable, and the small number of fire scenarios
which would impact the power supply to the motor-driven fire pump.
21EnclosureThe inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance (resources component) because procedure ST-O-37D-340-2 did not
provide complete and accurate instructions for cleaning the
- DD [[]]
FP cooling water strainer.
(IMC 0305 aspect
TS 5.4.1.a and NOM-C-7.1 require that procedures be used for equipmentmanipulations which could cause fire protection components to become inoperable.
Contrary to the above, procedures were not used when manipulating the
- DD [[]]
FP cooling
water isolation valves on May 23, 2007, resulting in the
- DD [[]]
FP being run on May 24, 2007,
without cooling water and sustaining engine damage. Because this failure to comply with
SRVsdid not meet their allowable leak rate for the pneumatic actuation controls for the
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. The corrective actions to resolve the underlying
causes of this event are in the
IR 539277). This licensee-identified violation was more than minor since it was associated with theEquipment Performance attribute of the Mitigating Systems Cornerstone and impacts the
cornerstone objective of ensuring the reliability, availability, and capability of systems that
respond to initiating events, in that, if the ADS system was called upon to actuate it's
operability would not be ensured. The inspectors evaluated this finding using IMC 0609,
Appendix A, "SDP of Reactor Inspector Findings for At-Power Situations," Phase 1
screening. Specifically, using the Mitigating Systems Cornerstone column, the inspectors
determined that a Phase 2 evaluation was required because the finding represented a
loss of system safety function. The inspectors concluded that the finding was of very low
safety significance (Green) because the success criteria for depressurization, on each of
the applicable worksheets, only required the use of 2 of 11 SRVs. A regional senior
reactor analyst reviewed and concurred with the inspectors risk assessment. This
licensee-identified finding involved a violation of TS 3.5.1, "Emergency Core Cooling
Systems." The enforcement aspects of this violation are discussed in Section 4OA7 of
this report. This LER is closed.
2Enclosure.5(Closed)
- LER 05000277/2006004-00, Plant Modification Created Diesel GeneratorBuilding Carbon Dioxide Suppression Room Flooding VulnerabilityOn November 17, 2006, engineering personnel determined that a potential floodvulnerability had existed in the
EDG building carbon dioxide suppression room. A plant
modification performed in 1985 had installed a catch basin at the EDG building fuel oil
filling station, which is located outside the EDG building. The catch basin discharge was
tied into the EDG building's oily waste separator tank, upstream of the flood protection
isolation valve. This constituted an unanalyzed condition that degraded plant safety. In
the event of a design basis flood, a potential pathway existed for flood water to enter the
building through the floor drains. It was determined that the maximum credible flow rate
would have exceeded the capability of the floor drain sump and sump pumps. Under
design basis flood conditions, the
- HP [[]]
SW system return valves would be challenged to perform their safety function.
Corrective actions recommended for this issue were documented in IR 554800 and
included revision of the applicable special event procedure for floods to mitigate this
condition. This finding is more than minor because it was associated with a degraded condition thatcould concurrently influence mitigation equipment. Specifically, with the degraded flood
barrier for the
ESW system booster
pumps and return valves and the
- HP [[]]
SW system return valves would be challenged to
perform their safety function under design basis flood conditions. The
- NRC [[]]
IMC 0609,
Appendix G, "Shutdown Operations SDP," applies because the plant would be shutdown,
at 112', in accordance with plant procedures, before flooding of EDG building would begin
to occur at the 128' elevation, as noted in the
LER, 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
2' elevation; flood alarms in the EDG building carbon dioxide suppression room that
would enable operators to take actions to stop the flooding; or operators could manually
operate the service water system return valves. A regional senior resident analyst
reviewed and concurred with the inspectors risk assessment. This licensee-identified
finding regarding the installation of a modification that placed the station in an unanalyzed
condition involved a violation of 10 CFR 50.59. The enforcement aspects of this violation
are discussed in Section 4OA7 of this report.
23Enclosure4OA5Other ActivitiesAs a plant status activity, the inspectors used guidance in
- NRC [[]]
IP 60855.1, "Operation ofan Independent Spent Fuel Storage Installation at Operating Plants," to selectively verify
that
- PBA [[]]
PS performed dry cask loading in a safe manner and in compliance with
approved procedures and work order instructions.4OA6Meetings, Including Exit.1Exit Meeting SummaryOn July 20, 2007, the resident inspectors presented the inspection results to Mr.
PBAPS staff, who acknowledged the findings. The inspectors
asked the licensee whether any of the material examined during the inspection should be
considered proprietary. No proprietary information was identified. .2Annual Assessment MeetingOn April 4, 2007, Mr. Paul Krohn, Mr. Mel Gray, the resident inspection staff, and otherNRC staff held a public meeting with Mr. Joe Grimes and other
- PBA [[]]
PS staff, to discuss
the results of the
PBAPS for the period January 1,
2006 through December 31, 2006. The handouts from the meeting are available
electronically from 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
VI of
the
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
II, Section E.2 b.1 states for State/Local Agencies: A
notification shall be made within fifteen (15) minutes of the initial emergency
classification. Contrary to this, on February 27, 2007, during an emergency event,
Peach Bottom personnel failed to notify one local county within 15 minutes of an
initial emergency declaration (Unusual Event); the notifications were completed in
minutes. The notification was not made in a timely manner because the
primary phone link to the county was not available. Plant procedures require the
notifications to be made using a backup phone. This finding is of very low safety
significance (Green) because the notification was late by only 3 minutes, backup
communication equipment was available, and procedures were available to use
the backup communication equipment. This was entered in
CAP as IR
596641.
24Enclosure*10 CFR 50.59, "Changes, Tests, and Experiments," requires, in part, that thelicensee may make changes in the facility as described in the safety analysis
report without prior Commission approval, unless the proposed change involves a
change in the TSs incorporated in the license or an unreviewed safety question
(USQ). A proposed change shall be deemed to involve a USQ, in part, if the
consequences of an accident or a malfunction of equipment important to safety
previously evaluated in the safety analysis report may be increased. Contrary to
this, in 1985, a change to the facility was made that remained in place until
November 2006, without analyzing whether a USQ existed. Specifically, as
documented in Section 4OA3.5, a plant modification performed in 1985 introduced
a potential flood vulnerability for the EDG building carbon dioxide suppression
room. The flood vulnerability posed by this change constituted an unanalyzed
condition that degraded plant safety. This was identified in
CAP as
IR 554800. This finding is of very low safety significance (Green) because the
likelihood of a design basis flood that could affect mitigation equipment is very
small and manual operator action could be taken to mitigate the effects of a design
basis flood.*TS 3.5.1, "Emergency Core Cooling Systems," requires that the
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
OA3.4, a
Phase 2 SDP determined that the finding was of very low safety significance
(Green) because the success criteria for depressurization, on each of the
applicable
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- INFORM [[]]
- POINTS [[]]
- OF [[]]
- CONTAC [[]]
TExelon Generation Company PersonnelJ. Grimes, Site Vice PresidentM. Massaro, Plant Manager
N. Alexakos, Manager, Engineering-Programs
J. Armstrong, Regulatory Assurance Manager
C. Behrend, Engineering Director
G. Jardel, Manager, Emergency Preparedness
C. Jordan, Chemistry Manager
D. Lewis, Operations Director
H. McCrory, Radiation Protection Technical Support Manager
M. Ross, Radwaste, Environmental Supervisor
G. Stathes, Maintenance Director
S. Taylor, Manager, Radiation Protection
NRC PersonnelF. Bower, Senior Resident InspectorM. Brown, Resident Inspector
R. Fuhrmeister, Senior Project Engineer
R. Nimitz, Senior Health Physicist
- LIST [[]]
- OF [[]]
- ITEMS [[]]
- AND [[]]
DISCUSSEDOpenedNone.Opened and Closed05000278/2007003-01FINInadequate Implementation of WOInstructions Caused the Installation of
an Incorrect Size Breaker and
Resulted in a Fire in the '4T4' 480 Volt
Load Center (Section
OA3.2)
A-2Attachment05000277, 278/2007003-03NCVInadequate Procedure AdherenceResults in Damage to the
- DD [[]]
URIIncorrect Size Breaker Resulted in aFire in the '4T4' 480 Volt Load Center
(Section
- LE [[]]
RPlant Modification Created DieselGenerator Building Carbon Dioxide
Suppression Room Flooding
Vulnerability (Section
- LIST [[]]
- OF [[]]
- REVIEW [[]]
- EDS ection 1R01: Adverse WeatherWC-AA-107, Revision 4, Seasonal ReadinessOP-AA-108-111-1001, Revision 2, Severe Weather and Natural Disaster Guidelines
- COL [[]]
HPSW System Startup and Normal Operatoins
A-3AttachmentP&ID DiagramM-315 Sheet 1, Revision 64,
HPSW Systems
M-315, Sheet 1, Revision 65,
HPSWSystemsM-330, Sheet 1, Revision 35, Emergency Cooling System
M-361, Sheet1, Revision 80, RHR System
M-361, Sheet 2, Revision 67,
- PF -70, Revision 2, Prefire Strategy Plan Standby Gas Treatment Room, Radwaste Building,91' 6" ElevationPF-13H, Revision 3, Prefire Strategy Plan North
- PF -55, Revision 3, Prefire Strategy Plan, Fire Zone 55, Unit 3 Refuel Floor, Reactor Building,234' ElevationPF-13D, Revision 1, Prefire Strategy Plan 3 'A' & 3 'C' Core Spray Rooms, Reactor Building,91'6" Elevation, Fire Zones 13D &
- PF -127, Revision 4, Prefire Strategy Plan, Unit 2 Emergency Battery/ Switchgear Room andRadwaste Corridor,
- TB -135PF-132, Revision 4, Prefire Strategy Plan, Diesel Generator Building, Elevation 127', FireZone 132PF-151, Revision 3, Prefire Strategy Plan, Unit 2 Main Transformer Yard, Fire Zone 151
- PF -164, Revision 0, Prefire Strategy Plan, 2 Startup Switchgear Building, Fire Zone 164Section 1R12: Maintenance Effectiveness
MCC Bucket Stabs
EHC Pump Breaker
ITE Rejection Tab Walkdown
IR 599203, Extent of Condition Walkdown of Unit 3 480 V Load Center Bus
ITE Breaker Found With No Rejection Tab
PCM Template Requirements
A-4AttachmentIR 623635,
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
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
- 3 CST [[]]
IR 623723, Bolt and Heli-coil Found Damaged at Disassembly on 00T634
SF-220, Revision 21, Spent Fuel Cask Loading and Transport Operations
A1406063, Review of Mod 79-028 Recirculation Seal Pressure Bleed Off
SO 2A.1.C-3, Revision 10, Operation of the Recirculation Pump Seal Purge System
A1439223,
ACMP - Unit 3, 3 'B' Recirculation Pump Seal Increasing Second Stage Seal Pressure
A1613202, 3 'B' Recirculation Pump 2nd Stage Seal PressureIR 619609, 3 'B' Recirculation Pump 2nd Stage Seal PressureARC 30C204M A-1, Revision 4 - A Recirculation Pump Seal Stage 2 Hi Flow
AA-101, "On-Line Work Control Process," Revision 13
SUE Outage Incomplete
IR 617946, Protected Equipment List Issued 4/16/07 Initially Incomplete
IR 504032, Exaggerated Paragon List of Protected Equipment
PRA Support for Protecting Equipment
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,
- SO 36A.1.A-2, Revision 2, Unit 2 'C' Air Compressor Return-to-Service and Service Air Systems Return to Normal Operation
- 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]]
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'
- CST [[]]
DH Sensing Line Leakage, datedMay 24, 2007A1615458, Small Leak on DPT-2-02-117D Line Snubber Threaded Cap
C0221439, Replace Snubber During an Outage
- PB [[]]
- ECR 03-00326 000, Revise Instrument Rack Drawings with a Note for SnubbersSection 1R19: Post-Maintenance TestingAR A1610537, High Lube Oil Temperature Alarm During E-2
- ODYN /B Minimum Critical Power Ratio (MCPR) RequirementsC0216504, PS-2-13-067-01: Replace Pressure Switch
RCIC Pump, Valve, Flow and Unit Cooler Functional andInservice Test, Conducted on April 5, 2007
A-6AttachmentC0215740,
MG Set and Power Distribution System Startup from Dead Bus ConditionR0629147, 3R4-U-C (7033B), Perform MCU Inspection
A1619582,
- PMT [[]]
- PMT [[]]
- SO 8G.6.A-3, Revision 3, Placing a Standby Off-Gas Glycol Pump in Service and Placing the InService Pump in Standby or OffSection 1R22: Surveillance TestingS12T-
- CRD Scram Insertion Timing of Selected Control Rods, Revision 19, completed May 5, 1997Section 1R23: Temporary Plant Modifications
- SP [[]]
- SO. 005-3, Revision 1, Routine Inspection of the 3 'A' Recirculation Seal Purge SupplementalCooling SystemSection
- ANS ) EvaluationPeach Bottom Nuclear Power Plant Upgraded Public Alert and Notification Report, April
- 2005FEMA [[]]
- 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
IRs:0043349400565056
00352078
005970650048176300451040
00520830
005213210053315700541478
00596641
A-7AttachmentSection
AA-113, Revision 7
06-16 "Radiological Emergency Plan"
AA-1000, Revision 17
06-33 "EP Plan Administration"
AA-120, Revision 7
06-96 "Emergency Preparedness Advisory Committee"
AA-120-1004, Revision 0
06-97 "Quarterly Satellite Phone Test"
MA-124-1004, Revision 0
06-99 "EP Fundamentals"
AA-1101, Revision 3
06-101"Exelon East
EP-MA-121-1002, Revision 4
06-102"Exelon East
EP-MA-121-1004, Revision 4
06-103"Exelon East
EP-MA-121-1005, Revision 3
06-108"ERO Fundamentals"
AA-1102, Revision 2
06-110"Mid-Atlantic
EP-AA-112-100-F-07
07-11"Exelon East
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
EP-AA-1007, Revision 14
07-39"Exelon East
- EP -AA-125 "Emergency Preparedness Self Evaluation Process," Revision4LS-AA-126 "Self-Assessment Program," Revision 5
AA-126-1001 "Focused Area Self-Assessments," Revision 4
Unusual Event Evaluation Reports dated 10/4/06, 11/21/06, 4/16/07
- AS [[]]
PEA-07-04 dated 5/9/07
A-8AttachmentIRs:00433494 00565056
00352078
004817630045104000520830
00533157
005414780059664100597065
00521321Section
- 1EP 6: Drill EvaluationPeach Bottom Atomic Power Station, May 15, 2007, Off-Year Exercise, Drill ScenarioPeach Bottom Atomic Power Station May 15th, 2007 Off-Year Exercise Report datedJune 14, 2007
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
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
- NRC [[]]
- 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
PI data, April 2006 - March 2007
Public Notification System
- DEP [[]]
- 4OA 2: Problem Identification and Resolution577381, Operator Failed to Perform Procedure Step581258, Page 12 of
ST-O-098-01N-2 Discovered Misplaced
568038,
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,
PB OPS Yellow For 4Q06
581376, Test Aborted: "ECT Portable Pump Operability" RT-O-48B-275-2
584506, Through Wall Leak Found on
ST-0-033-300-2 To Be Aborted
588335, Timeliness/Response to
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
593883, Unit 2 'C' RHR Sump Overflowed During Heat Exchange Maintenance
593890, Unit 2 'A' RHR Room Spill During Pumping of the Unit 2 'C' Room Sump
593891, Unit 2 'C' RHR Sump Overflowed During Heat Exchanger Maintenance
596641, Unusual Event Notification to York County Was > 15 Minutes
2264, Mid-Cycle Performance Gap - Self Assessment
606458, Training:
- PI [[]]
MS Code Improperly Granted
607064, Temperature Recorder TR-0558 not Functional (Discharge Canal)
615413, Non-Safety-Related Piece Part Installed in Diesel Generator
21191, Inadvertent
- PI [[]]
MS 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,
- DD [[]]
FP/ Engine Trip
604364, Human Error Results in E-3
- OA 3: Event FollowupSpecial Event Procedure (SE)-4, Flood, Revision 21 IR 563253, External Flood Vulnerability - Circulating Water Pump Structure
EDG Fire at Roof Exhaust Penetration
A-10AttachmentST-O-37D-370-2, Revision 25,
- DD [[]]
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
- LIST [[]]
- OF [[]]
- ACRONY MSADAMSAgency-wide Documents Access and Management SystemADSautomatic depressurization system
- AN [[]]
SAlert and Notification System
ARaction request
BTPsbranch technical positions
- CF [[]]
RCode of Federal Regulations
CRcondition report
CRDcontrol rod drive
CSTcondensate storage tank
- DD [[]]
- DE [[]]
- DO [[]]
- DR [[]]
PDivision of Reactor Projects
EALemergency action level
EDGemergency diesel generator
EPemergency preparedness
EROemergency response organization
- HP [[]]
- HP [[]]
SWhigh pressure service water
- IM [[]]
CInspection Manual Chapter
IPInspection Procedure
IRissue report
kVkilovolt
LERslicensee event reports
MRMaintenance Rule
MSmitigating system
- NE [[]]
- NR [[]]
CNuclear Regulatory Commission
OCsoperator challenges
OWAsoperator work-arounds
- PAR [[]]
- PBAP [[]]
- PC [[]]
- PC [[]]
PProcess Control Program
2AttachmentPIperformance indicatorPMTpost-maintenance testing
PROplant reactor operator
- RC [[]]
ICreactor core isolation cooling
RCRroot cause report
RFPreactor feed pump
RGRegulatory Guide
RHRresidual heat removal
RRPreactor recirculation pump
RTPrated thermal power
- SJ [[]]
- SP [[]]
ACstandards, policies, and administrative controls
SOsystem operating
SSCsstructures, systems, and components
SRVsafety relief valve
- TR [[]]
MTechnical Requirements Manual
TRTtroubleshooting, rework and testing
TSTechnical Specification
TSCtechnical support center
UEunusual event
- UFSA [[]]
RUpdated Final Safety Analysis Report
USQunreviewed safety question