IR 05000333/2007004
Download: ML073060352
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406 November 2, 2007 Mr. Peter Site Vice President Entergy Nuclear Northeast James A. FitzPatrick Nuclear Power Plant Post Office Box 110 Lycoming, NY 13093
SUBJECT: JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC INSPECTION REPORT 05000333/2007004
Dear Mr. Dietrich:
On September 30, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your James A. FitzPatrick Nuclear Power Plant. The enclosed inspection report documents the inspection results which were discussed on October 4, 2007 with Mr. K. Mulligan and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance 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. This report documents one finding of very low safety significance (Green). This finding did not involve a violation of NRC requirements.
In accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, 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 Web Site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/ Eugene W. Cobey, Chief Projects Branch 2 Division of Reactor Projects Docket No.: 50-333 License No.: DPR-59
Enclosure:
Inspection Report 05000333/2007004
w/Attachment:
Supplemental Information cc w/ encl: see next page
SUMMARY OF FINDINGS
.........................................................................................................iii
REPORT DETAILS
.....................................................................................................................1
REACTOR SAFETY
...................................................................................................................1 1R01 Adverse Weather Protection................................................................................1 1R04 Equipment Alignment...........................................................................................1 1R05 Fire Protection.....................................................................................................3 1R06 Flood Protection Measures..................................................................................3 1R11 Licensed Operator Requalification Program.........................................................4 1R12 Maintenance Effectiveness..................................................................................4 1R13 Maintenance Risk Assessments and Emergent Work Control..............................6 1R15 Operability Evaluations........................................................................................7 1R19 Post-Maintenance Testing...................................................................................8 1R20 Refueling and Other Outage Activities.................................................................8 1R22 Surveillance Testing............................................................................................9 1R23 Temporary Plant Modifications...........................................................................10 1EP6 Drill Evaluation...................................................................................................10
OTHER ACTIVITIES
................................................................................................................10
4OA1 Performance Indicator (PI) Verification..............................................................10 4OA2
Identification and Resolution of Problems..........................................................11
4OA3 Event Followup..................................................................................................12 4OA6
Meetings, Including Exit....................................................................................12 ATTACHMENT:
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-6
iiiSUMMARY
- OF [[]]
FINDINGS
IR 05000333/2007-004; 07/01/2007 - 09/30/2007; James A. FitzPatrick Nuclear Power Plant; Maintenance Effectiveness. The report covered a three-month period of inspection by resident inspectors. One Green
finding was 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
- 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 involving inadequate corrective actions when Entergy did not correct an adverse condition on the reactor core isolation cooling (RCIC) system flow instrument sensing lines. The condition allowed air
bubbles to form in the sensing lines, resulting in an erroneous flow indication. Consequently, the
- RCIC system would not have been able to achieve its design flow rate of 410 gallons per minute (gpm). Entergy entered the condition into their corrective action program and implemented interim corrective actions by revising the
RCIC operating procedure to vent the sensing lines. In addition, Entergy has
scheduled activities to correct the instrument sensing line condition. The inspectors determined that this finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone; and, it impacted the cornerstone objective of ensuring the availability,
reliability, and capability of the
- RCIC system would not have been able to achieve its design flow rate of 410 gpm. The inspectors evaluated this finding using Phase 1 of
- AS [[ignificance Determination of Reactor Inspection Findings for At-Power Situations,@ and determined it to be of very low safety significance (Green) because it was not associated with a design or qualification deficiency, it did not represent any actual loss of a system safety function, it did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, and it was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. (Section 1R12) B. Licensee-Identified Violations None.]]
- REPORT [[]]
DETAILS Summary of Plant Status The James A. FitzPatrick Nuclear Power Plant began the inspection period operating at full power. On August 20, 2007, Entergy shut down the plant to repair a leaking safety relief valve.
Following repairs, the plant was returned to full power on August 24, 2007. On September 12, 2007, operators initiated a manual reactor scram due to lowering plant cooling water intake level which was caused by lake algae intrusion. Following repairs to the traveling water screens and execution of a monitoring plan to assure availability of cooling water systems, the plant was started up on September 14, 2007, and returned to full power on September 16, 2007. The
plant continued to operate at or near full power for the remainder of the inspection period. 1.
SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather Protection (71111.01 - 1 sample) a. Inspection Scope The inspectors completed one adverse weather protection sample. High ambient
temperature and increased lake temperature were observed during early July 2007. To assess licensee actions and equipment performance, the inspectors reviewed Entergy's
actions and toured risk significant areas including the emergency diesel generator (EDG) building and east and west cable tunnels. The inspectors used Administrative Procedure 12.04, "Seasonal Weather Preparations," as a guide. The documents reviewed are listed in the Attachment. b. Findings No findings of significance were identified. 1R04 Equipment Alignment .1 Partial System Walkdown (71111.04Q - 4 samples) a. Inspection Scope The inspectors performed four partial system walkdowns to verify the operability of
redundant or diverse trains and components during periods of system train unavailability or following periods of maintenance. The inspectors referenced the system procedures, the Updated Final Safety Analysis Report (UFSAR), and system drawings in order to verify that the alignment of the available train was proper to support its required safety
2functions. The inspectors also reviewed applicable condition reports (CRs) and work orders to ensure that Entergy had identified and properly addressed equipment discrepancies that could potentially impair the capability of the available equipment train, as required by
XVI, "Corrective Action." The documents reviewed are listed in the Attachment. The inspectors performed a partial walkdown of the following systems which represented four inspection samples:
- 'A' low pressure coolant injection system when the 'B' low pressure coolant injection inverter was out of service for maintenance; * Reactor core isolation cooling system when the high pressure coolant injection system was out of service for testing; * High pressure coolant injection system when the reactor core isolation cooling system was out of service for testing; and * 'A' residual heat removal service water with 'B' residual heat removal service water out of service for planned maintenance. b. Findings No findings of significance were identified. .2 Complete System Walkdown (71111.04S - 1 sample) a. Inspection Scope The inspectors performed a complete system alignment inspection of the automatic depressurization system to identify any discrepancies between the existing equipment lineup and the required lineup. During the inspection, system drawings and operating procedures were used to verify proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders associated with the system for any deficiencies that could affect the ability of the system to perform its function. Documentation associated with unresolved design issues such as temporary
modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation. In addition, the inspectors reviewed the CR database to verify that the equipment problems were being identified and appropriately resolved. The documents reviewed are listed in the Attachment. The inspection represented one inspection sample. b. Findings No findings of significance were identified.
31R05 Fire Protection (71111.05Q - 9 samples) a. Inspection Scope The inspectors conducted a tour of several fire areas to assess the material condition and operational status of fire protection features. The inspectors verified, consistent with applicable administrative procedures, that: combustibles and ignition sources were adequately controlled; passive fire barriers, manual fire-fighting equipment, and
suppression and detection equipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program. The inspectors evaluated the fire protection program against the requirements of Licensee Condition 2.C.3. The documents reviewed are listed in the Attachment. This inspection
represented nine inspection samples for fire protection tours and was conducted in the following plant areas: * Fire Area/Zone
- SW [[-1, elevation 272 foot; * Fire Area/Zone Yard, elevation 272 foot; * Fire Area/Zone 1E/TB-1, elevation 300 foot; and * Fire Area/Zone 1E/TB-1, elevation 272 foot. b. Findings No findings of significance were identified. 1R06 Flood Protection Measures .1 Internal Flooding (71111.06 - 1 sample) a. Inspection Scope The inspectors reviewed selected risk-important plant design features and Entergy's procedures intended to protect the plant and its safety-related equipment from internal flooding events. The inspectors reviewed flood analysis and design documents, including the Individual Plant Examination and the]]
- UFS [[]]
AR, engineering calculations, and abnormal operating procedures. The documents reviewed are listed in the Attachment.
Inspections in the following plant areas represented one sample: * Relay room; * North cable tunnel; and * South cable tunnel.
b. Findings No findings of significance were identified. 1R11 Licensed Operator Requalification Program .1 Resident Inspector Quarterly Review (71111.11Q - 1 sample) a. Inspection Scope On August 13, 2007, the inspectors observed licensed operator simulator training to assess operator performance during several scenarios to verify that operator performance was adequate and evaluators were identifying and documenting crew
performance problems. The inspectors evaluated the performance of risk significant operator actions, including the use of emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation and manipulation, and the oversight and direction provided by the shift
manager. The inspectors also reviewed simulator fidelity to evaluate the degree of similarity to the actual control room. Licensed operator training was evaluated against the requirements of
- 10 CFR [[Part 55, "Operators' Licenses." The documents reviewed are listed in the Attachment. This observation of operator simulator training represented one inspection sample. b. Findings No findings of significance were identified. 1R12 Maintenance Effectiveness (71111.12Q - 3 samples) a. Inspection Scope The inspectors reviewed performance-based problems involving selected in-scope structures, systems, or components (]]
- SSC s) to assess the effectiveness of the maintenance program. The reviews focused on: * Proper Maintenance Rule scoping in accordance with
- 10 CFR Part 50.65; * Characterization of reliability issues; * Changing system and component unavailability; * 10
- CFR Part 50.65 (a)(1) and (a)(2) classifications; * Identifying and addressing common cause failures; * Trending of system flow and temperature values; * Appropriateness of performance criteria for
SSCs classified (a)(1).
5The inspectors reviewed system health reports, maintenance backlogs, and Maintenance Rule basis documents. The inspectors evaluated the maintenance program against the requirements of 10 CFR Part 50.65. The documents reviewed are listed in the Attachment. The following Maintenance Rule samples were reviewed and represent three inspection samples: * Reactor Core Isolation Cooling system; * Standby gas treatment system; and * Automatic depressurization system. b. Findings .1 Introduction: A Green self-revealing finding was identified involving inadequate corrective actions when Entergy did not correct an adverse condition on the reactor core
isolation cooling (RCIC) system flow instrument sensing lines. The condition allowed air bubbles to form in the sensing lines, resulting in an erroneous flow indication. Consequently, the
- RCIC system would not have been able to achieve its design flow rate of 410 gallons per minute (gpm). Description: On May 4, 2006, control room operators observed a reading of approximately 50 gpm on
RCIC flow instrumentation sensing lines, in conjunction with inadequate sloping of the instrument lines, led to the entrapment of air bubbles in
the sensing lines and the resultant erroneous flow indication. Entergy determined that
- RCIC injection flow rate is determined by an automatic flow controller, which adjusts the speed demand setting of the
- RC [[]]
IC turbine controls. The turbine speed is
determined by a comparison between the indicated flow as read on the
- RCIC [[flow controller adjusts the turbine speed demand to maintain indicated flow rate at the setpoint. Because indicated flow was higher than actual flow as a result of this condition, the actual injection flow rate would be less than desired. Entergy had previously identified problems with instrument line slope. On January 24, 2001, high pressure coolant injection (HPCI) system flow indicated 1400 gpm with the system in the standby condition.]]
- HPCI was declared inoperable and a root cause analysis was performed in which Entergy determined that the
- HPCI erroneous flow indication was caused by the negative sloping of the instrument sensing lines, air entrapment in the sensing lines and inadequate instrument venting. As part of the extent of condition review, the
- RCIC instrument lines could result in air entrapment and erroneous flow indications. Work Order (WO)
- HP [[]]
CI flow instrumentation in February,
2001. However, Equivalent Change
- RCIC instrument line slope to a minimum of one quarter inch per foot of sensing line run or +2.0 degrees, as required. Additionally, Entergy
- WO [[]]
- RCIC pump shutdown. Entergy entered the condition into their corrective action program and implemented interim corrective actions by revising the
- RC [[]]
IC operating procedure to vent the sensing
lines. In addition, Entergy has scheduled activities to correct the instrument sensing line slope. The inspectors determined that the performance deficiency was that Entergy did not correct the inadequate sloping of the
EN-LI-102, "Corrective Action Process," Revision 10, requires, in part, that corrective actions address the cause or resolve the deficiency. This was reasonably within Entergy=s ability to foresee and prevent. Traditional enforcement does not apply because the issue did not have an actual safety consequence or a potential for
impacting the
- NRC [[requirements. Analysis: The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone, and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of the]]
RCIC system would not have been able to achieve its
design flow rate of 410 gpm. The inspectors evaluated this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for At-Power Situations,@ and determined it to be of very low safety significance (Green) because it was not associated with a design or qualification deficiency, it did not represent any actual loss of a system safety function, it did not represent the actual loss of a safety function of a single train for greater than its
Technical Specification allowed outage time, and it was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. Enforcement: No violation of regulatory requirements occurred because corrective action issues related to the
FIN)05000333/2007004-01, Failure to Correct Negative Slope of the Reactor Core Isolation Cooling System Flow Instrument Sensing Lines.) 1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples) a. Inspection Scope The inspectors reviewed maintenance activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors
7verified that risk assessments were performed as required by
- 10 CFR [[Part 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The documents reviewed are listed in the Attachment. The review of the following activities represented five inspection samples. * Week of July 9, 2007, which included emergent work on 93 P1A1, 'A' emergency diesel generator (]]
EDG) fuel oil transfer pump; * Week of July 23, 2007, which included emergent work on 71T-4, normal station service transformer, to repair a terminal; * Week of August 13, 2007, which included emergent work on the 'B' low pressure coolant injection system inverter, emergent work on the torus vent valve, 27-AOV-118, and planned maintenance on the 'B' electro-hydraulic control pump; * Week of September 10, 2007, which included a manual reactor scram and emergent work on residual heat removal system strainers, normal service water strainers and intake screens; and * Week of September 17, 2007, which included EDG maintenance and emergent work on the 'A' emergency service water pump. b. Findings No findings of significance were identified. 1R15 Operability Evaluations (71111.15 - 5 samples) a. Inspection Scope The inspectors reviewed operability determinations to assess the acceptability of the evaluations; when needed, the use and control of compensatory measures; and
compliance with Technical Specifications (TS). The inspectors' review included a verification that the operability determinations were made as specified by
- OP -104, "Operability Determinations." The technical adequacy of the determinations was reviewed and compared to the
UFSAR, and associated design basis documents. The documents reviewed are listed in the Attachment. The following evaluations were
reviewed and represented five inspection samples: *
CR 2007-03202, concerning ultimate heat sink, emergency service water and residual heat removal service water fouling; and * CRs 2007-02711, and 2007-01236, concerning a cracked weld on the support structure of the 'A' standby gas treatment system fan.
b. Findings No findings of significance were identified. 1R19 Post-Maintenance Testing (71111.19 - 6 samples) a. Inspection Scope The inspectors reviewed six post-maintenance test procedures and associated testing
activities for selected risk-significant mitigating systems to assess whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel. The inspectors verified: test acceptance criteria were clear, demonstrated operational readiness, and were consistent with design basis documentation; test instrumentation had current calibrations and adequate range and
accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied. Upon completion, the inspectors verified that equipment was returned to the proper alignment necessary to perform its safety function. Post maintenance testing was evaluated against the requirements of
XI, "Test Control." The documents reviewed are listed in the
Attachment. The following post-maintenance test activities were reviewed and represented six inspection samples: * Work order 116184, involving repair of the 'A'
- EDG 02]]
RV-71F; * Work order 51104851, involving planned maintenance on the 'B' residual heat removal service water pump from September 10, 2007 through September 11, 2007; * Work order 00119231, involving the 'B' low pressure coolant injection inverter repair during week of August 12, 2007; and * Work order 00123589, involving emergent repairs to the 'A' emergency service water pump during the weeks of September 16, 2007 and September 23, 2007. b. Findings No findings of significance were identified. 1R20 Refueling and Other Outage Activities (71111.20 - 1 sample) a. Inspection Scope The inspectors observed and reviewed the following activities during the FitzPatrick scheduled maintenance outage from August 20, 2007 through August 22, 2007, to confirm that the Entergy had appropriately considered risk, industry experience, and previous site-specific problems in their outage plan. The documents reviewed are listed
in the Attachment. During the outage, the inspectors observed portions of the shutdown
9and cooldown and monitored licensee controls over the outage activities listed below. * The inspectors reviewed outage schedules and procedures and verified that Technical Specification required safety system availability was maintained, shutdown risk was considered, and that contingency plans existed to restore key safety functions such as electrical power and containment integrity. * The inspectors observed portions of the plant shutdown and cooldown and verified that the Technical Specification cooldown rate limits were not exceeded. * The inspectors periodically verified the proper alignment and operation of the shutdown cooling and reactor coolant makeup systems. * The inspectors observed portions of the reactor startup following the outage, and verified that safety-related equipment required for mode change was operable, containment integrity was set, and reactor coolant boundary leakage was within Technical Specification limits. b. Findings No findings of significance were identified. 1R22 Surveillance Testing (71111.22 - 4 samples) a. Inspection Scope The inspectors witnessed performance of surveillance tests and/or reviewed test data of
selected risk-significant
UFSAR, Technical Requirements Manual, and Entergy procedure requirements. The inspectors verified: test acceptance criteria were clear, demonstrated operational readiness, and were consistent with design basis documents; test instrumentation had current calibrations and adequate range and accuracy for the application; and tests were
performed, as written, with applicable prerequisites satisfied. Upon surveillance test (ST) completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. The inspectors evaluated the tests against the requirements in
STs were reviewed and represented four inspection samples:
CO2 Simulated Automatic/Manual Initiation Tests, South Emergency Switchgear Room." b. Findings No findings of significance were identified.
10 1R23 Temporary Plant Modifications (71111.23 - 1 sample) a. Inspection Scope The inspectors reviewed a temporary modification associated with the "A" drywell cooling coil leak repair conducted under
[[::JAF-2007-02926|JAF-2007-02926]]. The inspectors assessed the adequacy of the 10 CFR Part 50.59 evaluation for the temporary modification. The
inspectors also verified that the installation was consistent with the modification documentation; that the drawings and procedures were updated as applicable; and that the post-installation testing was adequate. The documents reviewed are listed in the Attachment. This review represented one inspection sample.
b. Findings No findings of significance were identified. 1EP6 Drill Evaluation (71114.06 - 1 sample) a. Inspection Scope The inspectors observed simulator activities associated with licensed operator requalification training on August 13, 2007. The inspectors verified that emergency classification declarations and notification activities were properly completed. The
inspectors evaluated the drill against the requirements of
- 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities." This observation represented one inspection sample. b. Findings No findings of significance were identified. 4.
- OTHER [[]]
- ACTIVI [[]]
- TIES [[]]
PI) Verification
a. Inspection Scope (71151 - 2 samples) The inspectors reviewed
- PI data for the cornerstone listed below and used Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guidance," Revision 5, to verify individual
PI accuracy and completeness.
11 Cornerstone: Barrier Integrity * Reactor coolant system leak rate * Reactor coolant system specific activity The inspectors reviewed operator logs, plant computer data, and surveillance procedure
- ST -40D, "Daily Surveillance and Channel Check," to verify the accuracy of Entergy's reported maximum reactor coolant system identified leakage for July 2006 to June 2007. b. Findings No findings of significance were identified. 4
OA2 Identification and Resolution of Problems .1 Review of Items Entered into the Corrective Action Program (CAP) a. Inspection Scope As required by Inspection Procedure 71152, "Identification and Resolution of Problems," and in order to help identify repetitive equipment failures or specific human performance
issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's
CAP items
across the initiating events, mitigating systems, and barrier integrity cornerstones for additional follow-up and review. The inspectors assessed Entergy's threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions. The CRs reviewed are listed in the Attachment. b. Findings and Observations No findings of significance were identified. .2 Annual Sample: Emergency Diesel Generator System (71152 - 1 sample)
a. Inspection Scope The inspectors selected several corrective action issues for detailed review that were associated with the emergency diesel generator (EDG) system. The main focus was on the maintenance, performance, and corrective actions associated with the EDGs after
the 'B'
CR-2007-1858. These reports were reviewed to ensure that an appropriate evaluation was performed and appropriate corrective actions were specified. The
2inspectors evaluated the reports against the requirements of procedure
- 10 CFR [[Part 50, Appendix B. b. Findings and Observations No findings of significance were identified. The inspectors determined that the causal analysis, extent of condition review, and the timeliness of the specified recommendations and corrective actions were appropriate. 4]]
OA3 Event Followup (71153 - 1 sample) .1 Manual Reactor Scram Due to Lowering Intake Water Level a. Inspection Scope The inspectors observed control room personnel responding to an unexpected decrease in plant cooling water intake level on September 12, 2007, which required that a manual reactor scram be initiated. As part of the followup to the event, the inspectors reviewed plant chart recorders, compared requirements of off-normal procedures to observations
of operators' performance, monitored equipment performance, and discussed the event response with plant personnel. The documents reviewed are listed in the Attachment. b. Findings No findings of significance were identified. 4OA6 Meetings, Including Exit On October 4, 2007, the inspectors presented the inspection results to Mr. Kevin J. Mulligan and other members of his staff. The inspectors asked Entergy
whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
- ATTACH [[]]
- MENT [[:]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- ATION [[]]
- KEY [[]]
- POINTS [[]]
- OF [[]]
CONTACT Entergy Personnel P. Dietrich, Site Vice President C. Adner, Manager Operations S. Bono, Director Engineering J. Costedio, Manager, Regulatory Compliance
- P. [[Cullinan, Manager, Emergency Preparedness M. Durr, Manager, System Engineering B. Finn, Director Nuclear Safety Assurance D. Johnson, Manager, Training J. LaPlante, Manager, Security K. Mulligan, General Manager, Plant Operations J. Pechacek, Manager, Programs and Components Engineering]]
- LIST [[]]
- OF [[]]
- ITEMS [[]]
- AND [[]]
DISCUSSED Opened None Opened and Closed 05000333/2007004-01 FIN Failure to correct negative slope of the reactor core isolation cooling system flow instrument sensing lines.
Closed None Discussed None
A-2LIST
- OF [[]]
- REVIEW [[]]
- 54 ER [[]]
PWR46, Fire Area/Zone 1E/TB-1
Section 1R06: Flood Protection Measures
RPT-MULTI-02107, "IPE Update,appendix H, Internal Flooding Analysis," Revision 2
Section 1R11: Licensed Operator Requalification Program
Eval 2006B, "RWR Pump Controller Failure Lowering Flow, Unisolable Torus Leak,
- AT [[]]
WS,
Degraded Emergency Depressurization"
A-3Section 1R12: Maintenance Effectiveness
RPT-NBS-12492, "Maintenance Rule Basis Document System 002, Nuclear Boiler, Automatic Depressurization, and Steam Leak Detection Systems," Revision 7; 02-ADS, "Automatic Depressurization System Health Report," Second Quarter 2007 Work Request 01-02670-00, "Re-slope line per JE-01-141"
Equivalent Change
- JAF -RPT-RCIC-02284, "Maintenance Rule Basis Document for System 013 Reactor Core Isolation Cooling System," Revision 4
- JAF -RPT-SGT-02495, "Maintenance Rule Basis Document for Systems 001-125 & 24 Standby Gas Treatment & Secondary Containment Systems," Revision
JAF Expert Panel Meeting Minutes, 7/26/05"
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
- FS [[]]
AR Figure No 5.3-2, "Standby Gas Treatment System," Revision 5
System Health Report for Standby Gas Treatment, 2nd Quarter 2007 Section 1R19: Post Maintenance Testing
- LPCI [[]]
- 50 DWG [[]]
A-4WO's 51101478 51101126 51193269 51101475 51100916
51193473 51101481 51100915 51193480 Section 1R20: Refueling and Other Outage Activities
- OP -13D, "RHR- Shutdown Cooling," Revision 18 AP-10.09, "Outage Risk Assessment," Revision 22 St-26J, "Heatup and Cooldown Temperature Checks," Revision 20
OP-109, "Drywell Leakage," Revision 1 AOP-39, "Loss of Coolant," , Revision 17
Section 1R23: Temporary Plant Modifications
[[::JAF-2007-02926|JAF-2007-02926]] Section 4OA1: Performance Indicator Verification
Procedure
LI-114,"Performance Indicator Process," Revision 2 Section 4OA2: Identification and Resolution of Problems
System Health Report for Emergency Diesel Generator System, 2nd Quarter 2007 System Health Report for Emergency Diesel Generator Ventilation, 2nd Quarter
- RPT -EDG-02303, "Maintenance Rule Basis Document for System 093 Emergency Diesel Generator System," Revision
RPT-DGV-02301, "Maintenance Rule Basis Document for System 092 Emergency Diesel
Generator Ventilation System," Revision 2
A-5 Condition Reports 2001-00308 2002-05008 2002-05107
2003-03214 2003-03278 2003-03308 2003-03318 2003-03532
2003-02911 2005-01004 2005-03157 2005-03468 2005-04453 2005-04487 2005-04843
2005-05255 2005-00772 2005-00795 2006-01726 2006-02890
2006-00596 2007-02503 2007-02854 2007-00752 2007-02050
2007-01530 2007-01236 2007-02711 2000-05406 2001-01923
2001-02194 2003-00678 2003-03768 2003-03717 2004-05270
2005-00178 2005-01468 2002-00231 2005-02844 2005-04113 2006-01989 2006-01846 2007-02147
2007-01827 2007-02306 2007-03363 2007-03263 2007-03266
2007-03267 2007-03268 2007-03269 2007-03270 2007-03273 2007-03361 2007-03370
2007-03259 2007-03260 2007-03148 2007-03154 2007-03155
2007-03158 2007-03160 2007-03162 2007-03166 2007-03167
2007-03133 2007-03135 2007-03138 2007-02931 2007-02945
2007-02899 2007-02901 2007-02905 2007-02906 2007-02907
2007-02908 2007-02909 2007-02911 2007-02912 2007-02916 2007-02917 2007-02918 2007-02921
2007-02925 2007-02926 2007-02876 2007-02825 2007-02823
2007-02821 2007-02956 2007-02958 2007-02964 2007-02966 2007-02968 2007-02972
2007-03335 2007-03347 2007-03350 2007-03352 2007-03354
2007-03358 2007-03359 2007-03361 2007-03321 2007-03331
2007-03332 2007-03333 2007-03300 2007-03302 2007-03304
2007-03315 2007-03078 2007-03083 2007-03086 2007-03087
2007-03281 2007-03284 2007-03289 2007-03293
Section
AOP-64, "Loss of Intake Water Level," Revision 7
A-6AOP-56, "High Traveling Screen or Trash Rack Differential Level," Revision 7 Transient 07-002, "Rx Scram Intake Blockage"
- LIST [[]]
- OF [[]]
- ACRONY [[]]
- MS [[]]