IR 05000333/2006005: Difference between revisions
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| issue date = 01/19/2007 | | issue date = 01/19/2007 | ||
| title = IR 05000333-06-005; 10/01/2006 - 12/31/2006; James A. FitzPatrick Nuclear Power Plant; Routine Resident Inspector Integrated Inspection Report | | title = IR 05000333-06-005; 10/01/2006 - 12/31/2006; James A. FitzPatrick Nuclear Power Plant; Routine Resident Inspector Integrated Inspection Report | ||
| author name = Cobey E | | author name = Cobey E | ||
| author affiliation = NRC/RGN-I/DRP/PB2 | | author affiliation = NRC/RGN-I/DRP/PB2 | ||
| addressee name = Dietrich P | | addressee name = Dietrich P | ||
| addressee affiliation = Entergy Nuclear Northeast | | addressee affiliation = Entergy Nuclear Northeast | ||
| docket = 05000333 | | docket = 05000333 | ||
| Line 18: | Line 18: | ||
=Text= | =Text= | ||
{{#Wiki_filter | {{#Wiki_filter:January 19, 2007Mr. Peter T. DietrichSite Vice President Entergy Nuclear Northeast James A. FitzPatrick Nuclear Power PlantPost Office Box 110 Lycoming, NY 13093SUBJECT:JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC INTEGRATEDINSPECTION REPORT 05000333/2006005 | ||
==Dear Mr. Dietrich:== | ==Dear Mr. Dietrich:== | ||
| Line 30: | Line 25: | ||
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.Based on the results of this inspection, no findings of significance were identified. However, alicensee-identified violation which was determined to be of very low safety significance is listed in the report. The NRC is treating this violation as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, because of the very low safety significance of the violation, and because it is entered into your corrective action program. If you contest the non-cited violation in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at the James A. FitzPatrick Nuclear Power Plant.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the P. Dietrich2NRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC'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). | The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.Based on the results of this inspection, no findings of significance were identified. However, alicensee-identified violation which was determined to be of very low safety significance is listed in the report. The NRC is treating this violation as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, because of the very low safety significance of the violation, and because it is entered into your corrective action program. If you contest the non-cited violation in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at the James A. FitzPatrick Nuclear Power Plant.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the P. Dietrich2NRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC'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, ChiefReactor Projects Branch 2 Division of Reactor ProjectsDocket No.:50-333License No.: DPR- | Sincerely, | ||
/RA/Eugene W. Cobey, ChiefReactor Projects Branch 2 Division of Reactor ProjectsDocket No.:50-333License No.: DPR-59Enclosure:Inspection Report 05000333/2006005w/ Attachment 1:Supplemental Information w/ Attachment 2: Mitigating System Performance Index Verificationcc w/encl:G. Taylor, CEO, Entergy Operations, Inc. | |||
Inspection Report 05000333/2006005w/ Attachment 1:Supplemental Information w/ Attachment 2: Mitigating System Performance Index Verificationcc w/encl:G. Taylor, CEO, Entergy Operations, Inc. | |||
M. Kansler, President, Entergy Nuclear Operations, Inc (ENO) | M. Kansler, President, Entergy Nuclear Operations, Inc (ENO) | ||
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==RADIATION SAFETY== | ==RADIATION SAFETY== | ||
===Cornerstone: | ===Cornerstone: Occupational Radiation Safety 2OS1Access Control to Radiologically Significant Areas (71121.01 - 7 samples) | ||
Occupational Radiation Safety 2OS1Access Control to Radiologically Significant Areas (71121.01 - 7 samples) | |||
====a. Inspection Scope==== | ====a. Inspection Scope==== | ||
During October 16 through 20, 2006, the inspector conducted the following activities toverify that Entergy was properly implementing physical, engineering, and administrativecontrols for access to high radiation areas (HRAs), and other radiologically controlledareas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the criteriacontained in 10 CFR 20, TS, and Entergy's procedures.*Radiation work permits (RWPs) were reviewed that provide access to exposuresignificant areas of the plant including HRAs. Specified electronic personaldosimeter alarm set points were reviewed with respect to current radiologicalcondition applicability and workers were queried to verify their understanding ofplant procedures governing alarm response and knowledge of radiologicalconditions in their work area.*There were no radiation work permits for airborne radioactivity areas with thepotential for individual worker internal exposures of greater than 50 millirem(mrem) committed effective dose equivalent (CEDE). | During October 16 through 20, 2006, the inspector conducted the following activities toverify that Entergy was properly implementing physical, engineering, and administrativecontrols for access to high radiation areas (HRAs), and other radiologically controlledareas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the criteriacontained in 10 CFR 20, TS, and Entergy's procedures.*Radiation work permits (RWPs) were reviewed that provide access to exposuresignificant areas of the plant including HRAs. Specified electronic personaldosimeter alarm set points were reviewed with respect to current radiologicalcondition applicability and workers were queried to verify their understanding ofplant procedures governing alarm response and knowledge of radiologicalconditions in their work area.*There were no radiation work permits for airborne radioactivity areas with thepotential for individual worker internal exposures of greater than 50 millirem(mrem) committed effective dose equivalent (CEDE).=== | ||
14Enclosure*The following radiologically significant work activities were selected; theradiological work activity job requirements were reviewed; and work activity jobperformance was reviewed with respect to the radiological work requirements.*In-Service inspection of reactor vessel nozzles;*Reactor vessel visual inspection and defueling activities;*Control rod drive replacement; and*High pressure turbine replacement.*During observation of the work activities listed above, the adequacy of surveys,job coverage and contamination controls were reviewed.*There were no significant dose gradients requiring relocation of dosimetry for theradiologically significant work activities listed above.*During observation of the work activities listed above, radiation workerperformance was evaluated with respect to the specific radiation protection (RP)work requirements and their knowledge of the radiological conditions in theirwork areas.*During observation of the work activities listed above, RP technician workperformance was evaluated with respect to their knowledge of the radiologicalconditions, the specific RP work requirements and RP procedures. | 14Enclosure*The following radiologically significant work activities were selected; theradiological work activity job requirements were reviewed; and work activity jobperformance was reviewed with respect to the radiological work requirements.*In-Service inspection of reactor vessel nozzles;*Reactor vessel visual inspection and defueling activities;*Control rod drive replacement; and*High pressure turbine replacement.*During observation of the work activities listed above, the adequacy of surveys,job coverage and contamination controls were reviewed.*There were no significant dose gradients requiring relocation of dosimetry for theradiologically significant work activities listed above.*During observation of the work activities listed above, radiation workerperformance was evaluated with respect to the specific radiation protection (RP)work requirements and their knowledge of the radiological conditions in theirwork areas.*During observation of the work activities listed above, RP technician workperformance was evaluated with respect to their knowledge of the radiologicalconditions, the specific RP work requirements and RP procedures. | ||
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(71151 - 7 samples)The inspectors reviewed performance indicator (PI) data for the below listedcornerstones and used NEI 99-02, "Regulatory Assessment PI Guidance," to verifyindividual PI accuracy and completeness.Cornerstone: Mitigating Systems*Safety system unavailability, RHR;*Safety system unavailability, HPCI;*Safety system unavailability, RCIC;*Safety system unavailability, Emergency AC Power; and*Safety system functional failures.The inspectors reviewed data and plant records from January 2004 to December 2006. The records reviewed included PI data summary reports, licensee event reports ,operator narrative logs, and maintenance rule records. The inspectors verified theaccuracy of the number of critical hours reported, and interviewed the system engineersand operators responsible for data collection and evaluation.Cornerstone: Occupational Radiation Safety*Occupational Exposure Control Effectiveness The inspector reviewed implementation of Entergy's Occupational Exposure ControlEffectiveness PI Program. Specifically, the inspector reviewed CRs, and radiologicalcontrolled area dosimeter exit logs for the past four calendar quarters. These recordswere reviewed for occurrences involving locked HRAs, very HRAs, and unplannedexposures against the criteria specified in Nuclear Energy Institute (NEI) 99-02, 16Enclosure"Regulatory Assessment PI Guideline", Revision 2, to verify that all occurrences that metthe NEI criteria were identified and reported as PIs. This inspection activity representsthe completion of one sample relative to this inspection area, completing the annualinspection requirement. | (71151 - 7 samples)The inspectors reviewed performance indicator (PI) data for the below listedcornerstones and used NEI 99-02, "Regulatory Assessment PI Guidance," to verifyindividual PI accuracy and completeness.Cornerstone: Mitigating Systems*Safety system unavailability, RHR;*Safety system unavailability, HPCI;*Safety system unavailability, RCIC;*Safety system unavailability, Emergency AC Power; and*Safety system functional failures.The inspectors reviewed data and plant records from January 2004 to December 2006. The records reviewed included PI data summary reports, licensee event reports ,operator narrative logs, and maintenance rule records. The inspectors verified theaccuracy of the number of critical hours reported, and interviewed the system engineersand operators responsible for data collection and evaluation.Cornerstone: Occupational Radiation Safety*Occupational Exposure Control Effectiveness The inspector reviewed implementation of Entergy's Occupational Exposure ControlEffectiveness PI Program. Specifically, the inspector reviewed CRs, and radiologicalcontrolled area dosimeter exit logs for the past four calendar quarters. These recordswere reviewed for occurrences involving locked HRAs, very HRAs, and unplannedexposures against the criteria specified in Nuclear Energy Institute (NEI) 99-02, 16Enclosure"Regulatory Assessment PI Guideline", Revision 2, to verify that all occurrences that metthe NEI criteria were identified and reported as PIs. This inspection activity representsthe completion of one sample relative to this inspection area, completing the annualinspection requirement. | ||
===Cornerstone: | ===Cornerstone: Public Radiation Safety* Radiological Environmental Technical Specifications/Offsite Dose CalculationManual- Radiological Effluent The inspector reviewed a listing of relevant effluent release reports for the past fourcalendar quarters, for issues related to the public radiation safety PI, which measuresradiological effluent release occurrences per site that exceed 1.5 millirem/quarter wholebody or 5.0 millirem/quarter organ dose for liquid effluents; 5 millirads/quarter gammaair dose, 10 millirad/quarter beta air dose, and 7.5 millirads/quarter for organ dose forgaseous effluents. This inspection activity represents the completion of one samplerelative to this inspection area, completing the annual inspection requirement. The inspector reviewed the following documents to ensure that Entergy met all of the PIrequirements:*Monthly projected dose assessment results due to radioactive liquid andgaseous effluent releases;*Quarterly projected dose assessment results due to radioactive liquid andgaseous effluent releases; and*Dose assessment procedures. | ||
Public Radiation Safety* Radiological Environmental Technical Specifications/Offsite Dose CalculationManual- Radiological Effluent The inspector reviewed a listing of relevant effluent release reports for the past fourcalendar quarters, for issues related to the public radiation safety PI, which measuresradiological effluent release occurrences per site that exceed 1.5 millirem/quarter wholebody or 5.0 millirem/quarter organ dose for liquid effluents; 5 millirads/quarter gammaair dose, 10 millirad/quarter beta air dose, and 7.5 millirads/quarter for organ dose forgaseous effluents. This inspection activity represents the completion of one samplerelative to this inspection area, completing the annual inspection requirement. The inspector reviewed the following documents to ensure that Entergy met all of the PIrequirements:*Monthly projected dose assessment results due to radioactive liquid andgaseous effluent releases;*Quarterly projected dose assessment results due to radioactive liquid andgaseous effluent releases; and*Dose assessment procedures. | |||
====b. Findings==== | ====b. Findings==== | ||
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====a. Inspection Scope==== | ====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 performanceissues for follow-up, the inspectors performed a daily screening of all items entered intoEntergy's corrective action program. The review was accomplished by accessingEntergy's computerized database for CRs and attending CR screening meetings. | As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"and in order to help identify repetitive equipment failures or specific human performanceissues for follow-up, the inspectors performed a daily screening of all items entered intoEntergy's corrective action program. The review was accomplished by accessingEntergy's computerized database for CRs and attending CR screening meetings.=== | ||
17EnclosureIn accordance with the baseline inspection modules, the inspectors selected 50corrective action program items across the Initiating Events, Mitigating Systems, andBarrier Integrity cornerstones for additional follow-up and review. | 17EnclosureIn accordance with the baseline inspection modules, the inspectors selected 50corrective action program items across the Initiating Events, Mitigating Systems, andBarrier Integrity cornerstones for additional follow-up and review. | ||
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===Closed=== | ===Closed=== | ||
05000333/2002002-01URIAdequacy of Hemyc Cable Wrap FireBarrier Qualification Test and Evaluation(Section 4OA5.2) | |||
==LIST OF DOCUMENTS REVIEWED== | ==LIST OF DOCUMENTS REVIEWED== | ||
==Section 1R02: Evaluation of Changes, Tests, or ExperimentsSafety EvaluationsJAF-SE-00-003, "Update of== | ==Section 1R02: Evaluation of Changes, Tests, or ExperimentsSafety EvaluationsJAF-SE-00-003, "Update of== | ||
: FSAR to Remove Inconsistency Concerning Maximum EDG Room Temperature," Revision 0JAF-SE-00-025, "Provide Chemical Cleaning Process for ESW Piping and Heat Exchanger," Revision 4JAF-SE-01-009, "Feedwater Heaters Maximum String Flow," Revision 0JAF-SE-05-001, "ASME Code Repair of Containment (Torus)," Revision 0JAF-SE-05-002, "Evaluation of the Capability of Cooling Water Intake Bar Heaters," Revision 0 | : FSAR to Remove Inconsistency Concerning Maximum EDG Room Temperature," Revision 0JAF-SE-00-025, "Provide Chemical Cleaning Process for ESW Piping and Heat Exchanger," Revision 4JAF-SE-01-009, "Feedwater Heaters Maximum String Flow," Revision 0JAF-SE-05-001, "ASME Code Repair of Containment (Torus)," Revision 0JAF-SE-05-002, "Evaluation of the Capability of Cooling Water Intake Bar Heaters," Revision 0 | ||
}} | }} | ||
Revision as of 06:33, 13 July 2019
| ML070190021 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/19/2007 |
| From: | Cobey E Reactor Projects Branch 2 |
| To: | Peter Dietrich Entergy Nuclear Northeast |
| cobey e w | |
| References | |
| IR-06-005 | |
| Download: ML070190021 (35) | |
Text
January 19, 2007Mr. Peter T. DietrichSite Vice President Entergy Nuclear Northeast James A. FitzPatrick Nuclear Power PlantPost Office Box 110 Lycoming, NY 13093SUBJECT:JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC INTEGRATEDINSPECTION REPORT 05000333/2006005
Dear Mr. Dietrich:
On December 31, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed aninspection at your James A. FitzPatrick Nuclear Power Plant. The enclosed inspection report documents the inspection results, which were discussed on January 4, 2007, with you and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.Based on the results of this inspection, no findings of significance were identified. However, alicensee-identified violation which was determined to be of very low safety significance is listed in the report. The NRC is treating this violation as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, because of the very low safety significance of the violation, and because it is entered into your corrective action program. If you contest the non-cited violation in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at the James A. FitzPatrick Nuclear Power Plant.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the P. Dietrich2NRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC'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, ChiefReactor Projects Branch 2 Division of Reactor ProjectsDocket No.:50-333License No.: DPR-59Enclosure:Inspection Report 05000333/2006005w/ Attachment 1:Supplemental Information w/ Attachment 2: Mitigating System Performance Index Verificationcc w/encl:G. Taylor, CEO, Entergy Operations, Inc.
M. Kansler, President, Entergy Nuclear Operations, Inc (ENO)
J. Herron, Sr, VP and Chief Operating Officer, (ENO)
C. Schwarz, VP, Operations Support (ENO)
K. Mulligan, General Manager, Plant Operations (ENO)
O. Limpias, VP, Engineering (ENO)
J. McCann, Director, Licensing (ENO)
C. Faison, Manager, Licensing (ENO)
M. Colomb, Director of Oversight (ENO)
D. Wallace, Director, Nuclear Safety Assurance (ENO)
J. Costedio, Manager, Regulatory Compliance (ENO)
T. McCullough, Assistant General Counsel (ENO)
P. Smith, President, New York State Energy Research and Development Authority P. Eddy, New York State Department of Public Service S. Lyman, Oswego County Administrator Supervisor, Town of Scriba C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law J. Sniezek, PWR SRC Consultant M. Lyster, PWR SRC Consultant S. Lousteau, Treasury Department, Entergy Services
SUMMARY OF FINDINGS
IR 05000333/2006-005; 10/01/2006 - 12/31/2006; James A. FitzPatrick Nuclear Power Plant;Routine Resident Inspector Integrated Inspection Report.The report covered a three-month period of inspection by resident inspectors, and announcedinspections by six regional specialist inspectors. The NRC's program for overseeing the safeoperation of commercial nuclear power reactors is described in NUREG-1649, "ReactorOversight Process," Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
B. Licensee-Identified Violations
A violation of very low safety significance, which was identified by Entergy, has beenreviewed by the inspectors. Corrective actions taken or planned by Entergy have beenentered into Entergy's corrective action program. The violation and corrective actiontracking number are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant StatusThe James A. FitzPatrick Nuclear Power Plant began the inspection period at 91 percent powerin a gradual power reduction (coastdown) as a result of fuel depletion at the end of theoperating cycle. On October 9, 2006, the plant was shutdown to commence a refueling outage. On November 3, 2006, reactor startup was commenced following completion of refuelingoutage activities. The generator was synchronized to the grid on November 5 and full powerwas achieved on November 7, 2006. The plant continued to operate at or near rated power forthe remainder of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 2 samples)
a. Inspection Scope
The inspectors completed the following two adverse weather protection samples.
- The inspectors reviewed Entergy's preparations for inclement weatherconditions, because high winds were forecast in the vicinity of the facility forDecember 2, 2006. The inspectors verified that operators implemented actionsand monitoring specified by Abnormal Operating Procedure (AOP)-13, "HighWinds, Hurricanes and Tornadoes;" and, toured the plant grounds and risksignificant areas including the screenwell, emergency diesel generator (EDG)building, and switchyard. In addition, the inspectors reviewed scheduledactivities for risk significant work that could be affected by high winds.*The inspectors reviewed and verified completion of the operations departmentcold weather preparation checklist contained in procedure AP-12.04, "SeasonalWeather Preparations," Revision 14. The inspectors reviewed the operatingstatus of the reactor and turbine building heating and cooling systems, EDGs,and fire protection water. Accessible areas of the reactor, turbine, and screenhouse buildings were inspected to assess the effectiveness of the ventilationsystems. The inspections included discussions with operations and engineeringpersonnel to ensure that they were aware of temperature restrictions andrequired actions. The inspectors also reviewed the following documents:*OP-51A, "Reactor Building Ventilation and Cooling System;"*OP-52, "Turbine Building Ventilation;"*DBD-066, "Design Basis Document for the Reactor Building Heating,Ventilation and Air Conditioning (HVAC) Systems;" and*DBD-067, "Design Basis Document for the Turbine Building HVACSystems."
b. Findings
No findings of significance were identified.
1R02 Evaluation of Changes, Tests, or Experiments (71111.02 - 17 samples) (6 safetyevaluations, 11 screenings)
a. Inspection Scope
The inspectors reviewed six safety evaluations in the Initiating Event, MitigatingSystems, and Barrier Integrity cornerstones. The selected safety evaluations werereviewed to verify that changes to the facility or procedures as described in the UpdatedFinal Safety Analysis Reports (UFSAR) were reviewed and documented in accordancewith 10 CFR 50.59, and that the safety issues pertinent to the changes were properlyresolved or adequately addressed. The inspectors assessed the adequacy of the safetyevaluations through interviews with the plant staff and review of supporting information,such as calculations and analyses, design change documentation, procedures, theUFSAR, Technical Specifications (TS) and plant drawings. The reviews also includedverification that Entergy had appropriately concluded that the changes and tests couldbe accomplished without obtaining license amendments.The inspectors also reviewed 11 evaluations for changes, tests and experiments forwhich Entergy determined that safety evaluations were not required. This review wasperformed to verify that Entergy's threshold for performing safety evaluations wasconsistent with 10 CFR 50.59.A listing of the evaluations reviewed is provided in Attachment 1.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04Q - 3 samples)Partial Walkdown
a. Inspection Scope
The inspectors performed three partial system walkdowns to verify the operability ofredundant or diverse trains and components during periods of system train unavailabilityor following periods of maintenance. The inspectors referenced the system procedures,the UFSAR, and system drawings in order to verify that the alignment of the availabletrain was proper to support its required safety functions. The inspectors also reviewedapplicable condition reports (CRs) and work orders to ensure that Entergy had identified 3Enclosureand properly addressed equipment discrepancies that could potentially impair thecapability of the available train, as required by 10 CFR Part 50, Appendix B, CriterionXVI, "Corrective Action." The documents reviewed are listed in Attachment 1. Theinspectors performed a partial walkdown on the following three systems:*Residual heat removal (RHR) system including containment spray inside thedrywell on October 18;*Reactor core isolation cooling system (RCIC) on November 6, following the highpressure coolant injection system failure on November 4; and*'B' and 'D' EDG subsystems on October 18, while subsystems 'A' and 'C' wereout of service for maintenance and testing.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05Q - 10 samples)Quarterly Inspection (10 - samples)
a. Inspection Scope
The inspectors conducted a tour of the ten areas listed below to assess the materialcondition and operational status of fire protection features. The inspectors verified thatcombustibles and ignition sources were controlled in accordance with Entergy'sadministrative procedures; fire detection and suppression equipment was available foruse; passive fire barriers were maintained; and compensatory measures forout-of-service, degraded, or inoperable fire protection equipment were implemented inaccordance with Entergy's fire plan. The inspectors used procedure ENN-DC-161,"Transient Combustible Program," in performing the inspection. The inspectorsevaluated the fire protection program against the requirements of License Condition2.C.3. The documents reviewed are listed in Attachment 1. This inspection satisfied teninspection samples for fire protection tours. The areas inspected included: *Fire Area/Zone XIV/PC-1, elevation 256 foot;*Fire Area/Zone XIV/PC-1, elevation 268 foot;*Fire Area/Zone IX/RB-1A, elevation 369 foot;*Fire Area/Zone IX/RB-1A, elevation 344 foot;*Fire Area/Zone IX/RB-1A, elevation 326 foot;*Fire Area/Zone X/RB-1;*Fire Area/Zone IX/SG-1; *Fire Area/Zone IA/MG-1;*Fire Area/Zone 1E/TB-1, elevation 252 foot; and*Fire Area/Zone 1E/OR-1.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures (71111.06 - 1 sample)Internal Flooding (1 sample)
a. Inspection Scope
The inspectors reviewed selected risk-important plant design features and licenseeprocedures intended to protect the emergency diesel generator buildings, cable tunnels, and associated safety-related equipment from internal flooding events. The inspectorsreviewed flood analysis and design documents, including the Individual PlantExamination (IPE) and the UFSAR, engineering calculations, and abnormal operatingprocedures. This inspection represented one sample.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the testing and evaluation of test results for the RHR systemheat exchangers performed in accordance with Entergy's response to Generic Letter(GL) 89-13, "Service Water System Problems Affecting Safety-Related Equipment."Heat removal measurements and heat exchanger capacity calculations were reviewedto verify that cooler performance was consistent with design calculations and theUFSAR. The documents reviewed are listed in the Attachment 1. This reviewrepresents one sample.
b. Findings
No findings of significance were identified.
5Enclosure1R08Inservice Inspection Activities (71111.08 - 1 sample)
a. Inspection Scope
The inspector observed non-destructive examination (NDE) activities and revieweddocumentation of NDE and repair activities. The sample selection was based on theinspection procedure objectives and risk priority of those components and systemswhere degradation could result in a significant increase in risk of core damage. Thedirect observations and documentation reviews were performed to verify that NDEactivities were performed in accordance with the American Society of MechanicalEngineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1989 Edition, noAddenda, 10CFR 50.55a, "Codes and Standards, Boiling Water Reactor VesselInternals Program" recommendations, and Entergy implementing procedures. Theinspector reviewed a sample of NDE reports initiated to document the performance andrecord results of in-service inspection (ISI) examinations completed during currentrefueling outage (RFO-17) as well as those since the last refueling outage. Theinspector also evaluated Entergy's effectiveness in resolving relevant indicationsidentified during ISI activities. Documents reviewed for this inspection are listed in theAttachment 1.The inspector reviewed several NDE examinations including visual, liquid penetrant,ultrasonic testing (UT), and radiographic examination data records to verify theeffectiveness of Entergy's program for monitoring degradation of risk significant pipingstructures, systems, and components. The inspector examined Entergy's evaluationand disposition for continued operation without repair or rework of non-conformingconditions identified during ISI activities by review of UT examination records NDEreport 06UT035 which was associated with a recordable indication in core spray weld12-14-734 and NDE report 06UT037 which evaluated the indication noted in NDE report06UT035. The report documented a relevant subsurface indication during UTexamination of an elbow to elbow weld in core spray system piping line 14-12"-W23-302-4A. Also, NDE report 04UT084, UT examination of weld 18-34-389 on feedwaterline 34-18"-WFP-902A, which had a recordable indication noted during UT examinationof a valve to elbow weld was considered a construction flaw, subsurface planerindication that was confirmed with radiographs.The inspector reviewed one ASME Section XI code repair and its associated NDE fromthe current refueling cycle. Specifically, the inspector reviewed welding repair activitiesand documentation performed on high pressure coolant injection (HPCI) outboardisolation 10" HPCI 23 MOV-16 valve to pipe weld #2 R1. This review was performed toverify that the activities associated with welding on ASME Class I or II components werein accordance with applicable ASME code requirements.The inspector performed direct field observations of the penetrant test (PT) examinationof standby liquid control nozzle 10 inside the drywell per work order [[::JAF-05-26173|JAF-05-26173]] and 6EnclosureUT examination of weld number 10-14-480 on core spray line 014-10" WD23-1504-5Ainside the drywell per work order [[::JAF-05-23900|JAF-05-23900]], including preparation and finalexamination data record review. The review was performed to evaluate examiner skillsand performance; examination technique; assess contractor oversight activities; and toverify Entergy's and the contractor's ability to identify and characterize observedindications.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
Resident Inspector Quarterly Review (71111.11Q - 1 sample)
a. Inspection Scope
On November 21, 2006, the inspectors observed licensed operator simulator training toassess operator performance during several scenarios to verify that operatorperformance was adequate and evaluators were identifying and documenting crewperformance problems. The inspectors evaluated the performance of risk significantoperator actions, including the use of emergency operating procedures. The inspectorsassessed the clarity and effectiveness of communications, the implementation ofappropriate actions in response to alarms, the performance of timely control boardoperation and manipulation, and the oversight and direction provided by the shiftmanager. The inspector also reviewed simulator fidelity with respect to the actual plant. Licensed operator training was evaluated against the requirements of 10 CFR 55,"Operators' Licenses." This observation of operator simulator training constituted one sample.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
(71111.12Q - 2 samples)
a. Inspection Scope
The inspectors reviewed performance-based problems involving selected in-scopestructures, systems, or components (SSCs) to assess the effectiveness of themaintenance program. Reviews focused on:*Proper Maintenance Rule (MR) scoping in accordance with 10 CFR 50.65;*Characterization of reliability issues; 7Enclosure* Changing system and component unavailability;* 10 CFR 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)(2); and
- Adequacy of goals and corrective actions for SSCs classified (a)(1).The inspectors reviewed system health reports, maintenance backlogs, andmaintenance rule basis documents. The inspectors evaluated the maintenance programagainst the requirements of 10 CFR 50.65. The documents reviewed are listed inAttachment 1. The following two maintenance rule samples were reviewed:*Service water pumps and area unit coolers; and*Reactor building closed loop cooling water system containment isolation valves.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
(71111.13 - 5 samples)
a. Inspection Scope
The inspectors reviewed the following five activities to verify that the appropriate riskassessments were performed prior to removing equipment for work. The inspectorsverified that risk assessments were performed as required by 10 CFR 50.65(a)(4), andwere accurate and complete. When emergent work was performed, the inspectorsverified that the plant risk was promptly reassessed and managed. The documentsreviewed are listed in Attachment 1. The following activities represent five inspection samples:*Week of October 2, 2006, that included emergent work on the east diesel-drivenfire pump, one 115 kilovolt (kV) offsite circuit inoperable for maintenance, andpre-outage surveillance activities;*Week of October 16, 2006, that included planned maintenance on both trains ofthe standby gas treatment system;*Week of November 6, 2006, that included emergent work on the 120 VoltAlternating Current uninterruptible power supply;*Week of November 13, 2006, that included replacement of 24 Volt Direct Currentinstrument batteries; and*Week of December 18, 2006, that included emergent work on the fire systempressure maintenance pump and inoperable fire protection hydrants during fireprotection system troubleshooting.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15 - 6 samples)
a. Inspection Scope
The inspectors reviewed operability determinations to assess the acceptability of theevaluations; when needed, the use and control of compensatory measures; andcompliance with TSs. The inspectors' review included a verification that the operabilitydeterminations were made as specified by ENN-OP-104, "Operability Determinations." The technical adequacy of the determinations was reviewed and compared to the TS,the UFSAR, and associated design basis documents. The documents reviewed arelisted in the Attachment 1. The following six evaluations were reviewed and eachconstituted inspection program samples:*CR-2006-04484 concerning identification of a crack in the torus wall at theinternal torus gusset plate weld;*CRs-2006-04224, 04227, and 04183 concerning shroud ring segmentindications;*CR-2006-04327 concerning less than minimum required thrust on low pressurecoolant injection valve 10MOV-25A;*CR-2006-04294 concerning less than minimum required thrust on HPCI steamsupply containment isolation valve 23MOV-15;*CR-2006-04739 concerning low individual cell voltages in the 'A' low pressurecoolant injection system battery; and*CR-2006-04941 concerning the inoperability of control room chiller 70RWC-2B.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications (71111.17A - 2 samples, 71111.17B - 6 samples).1Annual Inspection (2 - samples)
a. Inspection Scope
The inspectors reviewed design and post-installation test documents for modificationsER-JAF-05-25576, which installed a sparger on the HPCI turbine steam exhaust pipingin the torus, and ER-JAF-05-19832, which replaced the HPCI steam supply outboardcontainment isolation valve 23MOV-16. The HPCI steam exhaust sparger was installedto eliminate the effects of condensation oscillation loads on the torus shell that had 9Enclosureresulted in through-wall cracks. Original double disk gate valve 23MOV-16 wasreplaced with a split wedge gate design to improve local leak rate test performance. Post-installation tests of the HPCI system confirmed sparger design assumptions. Valve 23MOV-16 tests included valve diagnostic, inservice, and seat leakage tests, andnondestructive examinations. Documents reviewed for this inspection are listed in theAttachment 1.
b. Findings
No findings of significance were identified..2Biennial Inspection (71111.17B - 6 samples)The inspectors reviewed six risk-significant plant modification packages. The reviewwas performed to verify that the design bases, licensing bases, and performancecapability of risk significant SSCs had not been degraded through the modifications. The selected plant modifications were distributed among the Initiating Event, MitigatingSystems, and Barrier Integrity cornerstones. For the accessible components associatedwith the modifications, the inspectors walked down the systems to detect possibleabnormal installation conditions. The inspectors verified that selected attributes wereconsistent with the design and licensing bases. These attributes included componentsafety classification, energy requirements supplied by supporting systems, instrumentsetpoints, and supporting electrical and mechanical calculations and analyses. Designassumptions were reviewed to verify that they were technically appropriate andconsistent with the UFSAR. For selected permanent plant changes, the 50.59 screensor evaluations were reviewed as described in section 1R02 of this report. Theinspectors verified that procedures, calculations and the UFSAR were properly updatedwith revised design information and operating guidance. The inspectors also verifiedthat post-modification testing was adequate to ensure the SSC would function inaccordance with its design assumptions. The modifications reviewed are listed in theAttachment 1.
b. Findings
No findings of significance were identified.
10Enclosure1R19Post Maintenance Testing (71111.19 - 9 samples)
a. Inspection Scope
The inspectors reviewed nine post maintenance test procedures and associated testingactivities for selected risk significant mitigating systems to assess whether the effect ofmaintenance on plant systems was adequately addressed by control room andengineering personnel. The inspectors verified that test acceptance criteria were clear,demonstrated operational readiness and were consistent with design basisdocumentation; test instrumentation had current calibrations and the range andaccuracy for the application; and tests were performed, as written, with applicableprerequisites satisfied. Upon completion, the inspectors verified that equipment wasreturned to the proper alignment necessary to perform its safety function. Postmaintenance testing was evaluated against the requirements of 10 CFR 50, Appendix BCriterion XI, "Test Control." The following nine post maintenance test activities werereviewed and represent nine inspection program samples.*Work Request (WR) [[::JAF-06-28641|JAF-06-28641]] and WR [[::JAF-06-28820|JAF-06-28820]] involved excavationand weld repairs of two torus upper ring girder gusset-to-shell welds. Theretests consisted of UT measurements and pneumatic leakage tests of therepaired areas in accordance with ASME Boiler and Pressure Vessel CodeSection XI, 1992 edition, Articles IWE-2500 and 5221. *WR [[::JAF-05-31815|JAF-05-31815]] involved replacement of containment instrument nitrogenaccumulator check valves and tubing connections to the safety relief valves. Theretest consisted of a leak rate test using ST-39B, "Leak Rate Test of ADSPneumatic Supply Check Valves."*WR [[::JAF-06-30367|JAF-06-30367]] involved replacing 'A' train low pressure coolant injectionsystem battery cells. The retest consisted of verifying individual cell voltagesand specific gravities using MST-071.11, "LPCI Battery Quarterly SurveillanceTest."*WR [[::JAF-06-28896|JAF-06-28896]] and WR [[::JAF-06-28937|JAF-06-28937]] involved repair of drywell equipmentdrain pump discharge containment isolation valve 20AOV-95 following a failedlocal leak rate test. The retests consisted of a local leak rate test, valveexercising, and valve position indication verification using ST-39B-X19, "Type CLeak Test of Drywell Equipment Drain Discharge Valves (IST)," ST-1C, "PrimaryContainment Isolation Valve Exercise Test (IST)," and ST-41D, "Remote ValvePosition Indication Verification," respectively.*WR [[::JAF-06-25666|JAF-06-25666]] and WR [[::JAF-05-34791|JAF-05-34791]] involved replacing 'B' train lowpressure coolant injection system battery cells. The retest consisted of verifyingindividual cell voltages and specific gravities using MST-071.11 "Low PressureCoolant Injection Battery Quarterly Surveillance Test."*WR [[::JAF-06-32359|JAF-06-32359]] involved repair of the switch trip arm on containment isolationvalve 20AOV-95. The retest consisted of a valve exercise test and verification ofremote position indication using ST-1C, "Primary Containment Isolation Valve 11EnclosureExercise Test (IST)," and ST-41D, "Remote Valve Position IndicationVerification."*WR [[::JAF-06-32136|JAF-06-32136]] involved replacement of the fire protection system pressuremaintenance pump. The retest consisted of performance of ST-76D, "HighPressure Water Fire Protection System Operational Test."*WR [[::JAF-05-36422|JAF-05-36422]] involved repair of the 'D' main steam isolation valve. Theretest consisted of performance of ST-39B-X7D, "Type C Leak Test Main SteamLine D Main Steam Isolation Valves."*WR [[::JAF-06-26857|JAF-06-26857]] involved intermediate range monitor (IRM) 'D' repair. Theretest consisted of performance of ISP-71-1, "Intermediate Range Monitor LogicSystem Functional Test."
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 selected refueling outage activities to verify that operability requirements were met and that risk, industry experience, and previous sitespecific problems were considered.*Outage Plan: The inspectors reviewed outage schedules and procedures, andverified that TS-required safety system availability was maintained, shutdownrisk was considered, and contingency plans existed for restoring key safetyfunctions such as electrical power and primary coolant system makeup. *Plant shutdown and cooldown: The inspectors observed portions of the plantshutdown and cooldown on October 8 and 9, and verified that the TS cooldownrate limits were satisfied.*During the course of the refueling outage, the inspectors observed selectedreactor disassembly activities and walked down clearances to verify that tagoutswere properly hung and that equipment was properly configured. Through planttours, the inspectors verified that Entergy maintained and adequately protectedelectrical power supplies to safety-related equipment and that TS requirementswere met.*The inspectors periodically verified proper alignment and operation of theshutdown cooling and alternate decay heat removal systems. The verificationalso included reactor cavity and fuel pool makeup paths and water sources, andadministrative control of drain down paths.
12Enclosure*The inspectors reviewed procedures RAP-7.1.04B, "Refueling Procedure," andRAP-7.1.04C, "Neutron Instrument Monitoring During In-Core Fuel Handling,"and the results of refueling platform interlock functional tests to ensure that theTS requirements for fuel movement were met. The inspectors also verifiedthrough review of procedure ST-39D, "Secondary Containment Leak Test," thatcontainment requirements for refueling activities were met.*The inspectors observed portions of the reactor startup following the outage, andverified through plant walkdowns, control room observations, and surveillancetest reviews that the safety-related equipment required for mode change wasoperable, that containment integrity was set, and that reactor coolant boundaryleakage was within TS limits.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22 - 6 samples)
a. Inspection Scope
The inspectors witnessed performance of surveillance tests and/or reviewed test data ofselected risk-significant SSCs to assess whether the SSCs satisfied TSs, UFSAR,technical requirements manual, and Entergy procedure requirements. The inspectorsverified that test acceptance criteria were clear, demonstrated operational readiness andwere consistent with design basis documentation; that test instrumentation had currentcalibrations and the range and accuracy for the application; and that tests wereperformed, as written, with applicable prerequisites satisfied. Upon completion, theinspectors verified that equipment was returned to the status specified to perform itssafety function. The inspectors evaluated the surveillance tests against the requirementsin TS. The following surveillance tests were reviewed and represented six inspectionprogram samples:*ST-9C, "Emergency AC Power Load Sequencing Test and 4 kV EmergencyPower System Voltage Relays Instrument Functional Test;" *ST-39B-X9A/B, "Type C Leak Test of Feed Water System Line 'A' and 'B' Valves(IST);"*ST-39H, "Reactor Pressure Vessel Leakage Test;"*ST-9BA, "EDG 'A' and 'C' Full Load Test and Emergency Service Water PumpOperability Test;"*ST-6M, "Standby Liquid Control Recirculation, Injection Test (IST, ISI);" and *MST-071.26, "Station Battery 'A' Modified Performance Test."
b. Findings
No findings of significance were identified.Cornerstone: Emergency Preparedness
[EP]1EP6Drill Evaluation (71114.06 - 1 sample)
a. Inspection Scope
The inspectors observed simulator activities associated with licensed operatorrequalification training on November 21. The inspectors verified that emergencyclassification declarations and notification activities were properly completed.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
===Cornerstone: Occupational Radiation Safety 2OS1Access Control to Radiologically Significant Areas (71121.01 - 7 samples)
a. Inspection Scope
During October 16 through 20, 2006, the inspector conducted the following activities toverify that Entergy was properly implementing physical, engineering, and administrativecontrols for access to high radiation areas (HRAs), and other radiologically controlledareas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the criteriacontained in 10 CFR 20, TS, and Entergy's procedures.*Radiation work permits (RWPs) were reviewed that provide access to exposuresignificant areas of the plant including HRAs. Specified electronic personaldosimeter alarm set points were reviewed with respect to current radiologicalcondition applicability and workers were queried to verify their understanding ofplant procedures governing alarm response and knowledge of radiologicalconditions in their work area.*There were no radiation work permits for airborne radioactivity areas with thepotential for individual worker internal exposures of greater than 50 millirem(mrem) committed effective dose equivalent (CEDE).===
14Enclosure*The following radiologically significant work activities were selected; theradiological work activity job requirements were reviewed; and work activity jobperformance was reviewed with respect to the radiological work requirements.*In-Service inspection of reactor vessel nozzles;*Reactor vessel visual inspection and defueling activities;*Control rod drive replacement; and*High pressure turbine replacement.*During observation of the work activities listed above, the adequacy of surveys,job coverage and contamination controls were reviewed.*There were no significant dose gradients requiring relocation of dosimetry for theradiologically significant work activities listed above.*During observation of the work activities listed above, radiation workerperformance was evaluated with respect to the specific radiation protection (RP)work requirements and their knowledge of the radiological conditions in theirwork areas.*During observation of the work activities listed above, RP technician workperformance was evaluated with respect to their knowledge of the radiologicalconditions, the specific RP work requirements and RP procedures.
b. Findings
No findings of significance were identified.2OS2ALARA Planning and Controls (71121.02 - 3 samples)
a. Inspection Scope
During October 16 through 20, 2006, the inspector conducted the following activities toverify that Entergy was properly maintaining individual and collective radiation exposuresas low as is reasonably achievable (ALARA). Implementation of the ALARA programwas reviewed against the criteria contained in 10 CFR 20.1101(b) and Entergy'sprocedures.*The following highest exposure work activities for the Fall 2006 refueling outagewere selected for review:*In-Service inspection of reactor vessel nozzles;*Reactor vessel visual inspection and defueling activities;*Control rod drive replacement; and*High pressure turbine replacement.
15Enclosure*With respect to the work activities listed above, these job sites were observed toevaluate if surveys and ALARA controls were implemented as planned.*With respect to the work activities listed above, radiation worker and RPtechnician performance was observed during the performance of these workactivities to demonstrate the ALARA principles.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1Performance Indicator Verification
a. Inspection Scope
(71151 - 7 samples)The inspectors reviewed performance indicator (PI) data for the below listedcornerstones and used NEI 99-02, "Regulatory Assessment PI Guidance," to verifyindividual PI accuracy and completeness.Cornerstone: Mitigating Systems*Safety system unavailability, RHR;*Safety system unavailability, HPCI;*Safety system unavailability, RCIC;*Safety system unavailability, Emergency AC Power; and*Safety system functional failures.The inspectors reviewed data and plant records from January 2004 to December 2006. The records reviewed included PI data summary reports, licensee event reports ,operator narrative logs, and maintenance rule records. The inspectors verified theaccuracy of the number of critical hours reported, and interviewed the system engineersand operators responsible for data collection and evaluation.Cornerstone: Occupational Radiation Safety*Occupational Exposure Control Effectiveness The inspector reviewed implementation of Entergy's Occupational Exposure ControlEffectiveness PI Program. Specifically, the inspector reviewed CRs, and radiologicalcontrolled area dosimeter exit logs for the past four calendar quarters. These recordswere reviewed for occurrences involving locked HRAs, very HRAs, and unplannedexposures against the criteria specified in Nuclear Energy Institute (NEI) 99-02, 16Enclosure"Regulatory Assessment PI Guideline", Revision 2, to verify that all occurrences that metthe NEI criteria were identified and reported as PIs. This inspection activity representsthe completion of one sample relative to this inspection area, completing the annualinspection requirement.
===Cornerstone: Public Radiation Safety* Radiological Environmental Technical Specifications/Offsite Dose CalculationManual- Radiological Effluent The inspector reviewed a listing of relevant effluent release reports for the past fourcalendar quarters, for issues related to the public radiation safety PI, which measuresradiological effluent release occurrences per site that exceed 1.5 millirem/quarter wholebody or 5.0 millirem/quarter organ dose for liquid effluents; 5 millirads/quarter gammaair dose, 10 millirad/quarter beta air dose, and 7.5 millirads/quarter for organ dose forgaseous effluents. This inspection activity represents the completion of one samplerelative to this inspection area, completing the annual inspection requirement. The inspector reviewed the following documents to ensure that Entergy met all of the PIrequirements:*Monthly projected dose assessment results due to radioactive liquid andgaseous effluent releases;*Quarterly projected dose assessment results due to radioactive liquid andgaseous effluent releases; and*Dose assessment procedures.
b. Findings
No findings of significance were identified.4OA2Identification and Resolution of Problems.1Review 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 performanceissues for follow-up, the inspectors performed a daily screening of all items entered intoEntergy's corrective action program. The review was accomplished by accessingEntergy's computerized database for CRs and attending CR screening meetings.===
17EnclosureIn accordance with the baseline inspection modules, the inspectors selected 50corrective action program items across the Initiating Events, Mitigating Systems, andBarrier Integrity cornerstones for additional follow-up and review.
Additionally, NRCspecialist inspectors reviewed CRs associated with ISI, occupational radiation safetyand 10 CFR 50.59 plant modifications. The inspectors assessed Entergy's threshold forproblem identification, the adequacy of the cause analyses, extent of condition review,and operability determinations, and the timeliness of the specified corrective actions. The CRs reviewed are noted in the Attachment 1. b.Assessment and Observations No findings of significance were identified.
.2 Semi-Annual Review to Identify Trends
a. Inspection Scope
As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"the inspectors performed a review of Entergy's Corrective Action Program andassociated documents to identify trends that could indicate the existence of a moresignificant safety issue. The inspectors' review was focused on repetitive equipmentand corrective maintenance issues but also considered the results of daily inspectorCAP item screening discussed in Section 4OA2.1. The review also included issuesdocumented outside the normal CAP in system health reports, corrective maintenancework requests, component status reports, site monthly meeting reports andmaintenance rule assessments. The inspectors' review nominally considered thesix-month period of July 2006 through December 2006, although some examplesexpanded beyond those dates when the scope of the trend warranted. The inspectorscompared and contrasted their results with the results contained in the licensee's latestintegrated quarterly assessment report. Corrective actions associated with a sample ofthe issues identified in the licensee's trend report were reviewed for adequacy. Theinspectors also evaluated the trend report specified in ENN-LI-102, "Corrective ActionProcess," and 10 CFR 50, Appendix B. The documents reviewed during this inspectionare listed in the Attachment 1. b.Assessment and ObservationsEquipment, human performance and program issues were identified at an appropriatethreshold and were entered into their corrective action program. No findings ofsignificance were identified.
18Enclosure.3Annual PI&R Sample Review (71152 - 3 samples)
a. Inspection Scope
The inspectors selected three corrective action issues for detailed review. Thesereports were reviewed to ensure that an appropriate causal evaluation was performedand appropriate corrective actions were specified. The inspectors evaluated the reportsagainst the requirements of procedure ENN-LI-102, "Corrective Action Process," and10 CFR 50, Appendix B. Documents reviewed are listed in the Attachment 1.* CR-2006-03286 concerned the loss of the 'B' reactor protection system (RPS)bus and half scram that occurred on September 8, 2006;*CRs 2006-03543, 05149, 04491, and 04778 involved relief valve inservice testfailures in the emergency service water and shutdown cooling systems; and*CR 2006-01044 involved 'B' reactor feed pump mechanical seal leakage. b.Assessment and ObservationsNo findings of significance were identified. The adequacy of causal analysis, extent ofcondition review and the timeliness of the specified corrective actions were determinedto be reasonable.4OA5Other Activities.1Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review
a. Inspection Scope
In accordance with the NRC Field Policy Manual, NUREG/BR-0075, Revision 4, theinspectors reviewed the final report for the INPO plant assessment of FitzPatrickconducted in March, 2006. The inspectors reviewed the report to ensure that issuesidentified were consistent with the NRC perspectives of licensee performance and toverify if any significant safety issues were identified that required further NRC follow-up.
b. Findings
No findings of significance were identified..2(Closed) URI 05000333/2002002-01, Adequacy of Hemyc Cable Wrap Fire BarrierQualification Test and Evaluation Inspection report 05000333/2002002 documented the potential inadequacy of Hemycfire wrap barrier material at FitzPatrick. The issue was unresolved pending further NRC 19Enclosurereview to determine whether the qualification tests of the Hemyc fire wrap systems wereacceptable. In subsequent NRC fire tests, results indicated that Hemyc/MT materialscannot be routinely relied upon as one hour fire barriers. The NRC staff has completeda significant effort informing industry of the concerns associated with these materials byissuing Information Notice 2005-07, "Results of Hemyc Electrical Raceway Fire BarrierSystem Full Scale Fire Testing," and Generic Letter (GL) 2006-03, "PotentiallyNonconforming Hemyc and MT Fire Barrier Configurations." As required by GL2006-03, Entergy responded appropriately to the NRC concerns by identifying allapplications of Hemyc/MT materials, implementing compensatory measures, andinitiating corrective actions. On September 27, 2006, the NRC approved an exemptionfrom 10 CFR 50, Appendix R pertaining to the Hemyc fire wrap installed in the westcable tunnel. The NRC staff has determined that there was no performance deficiencyassociated with the issue and this unresolved item is closed..3Temporary Instruction (TI) - 2515/169, "Mitigating System Performance IndexVerification"
a. Inspection Scope
The objective of TI 2515/169 is to verify that the licensee has correctly implemented theMitigating Systems Performance Index (MSPI) guidance for voluntarily reportingunavailability and unreliability of the monitored safety systems. On a sampling basis,the inspector validated the accuracy of the unavailability and unreliability input data usedfor both the 12-quarter period of baseline performance and for the first reported results(second calendar quarter 2006). Specific attributes examined by the inspectors per thisTI included: surveillance activities which, when performed, do not render the trainunavailable for greater than 15 minutes; surveillance activities which, when performed,do not render the train unavailable due to credit for prompt operator recovery actions;and for each MSPI system, on a sampling basis, the inspectors independently confirmedthe accuracy of baseline planned unavailability, actual planned and unplannedunavailability, and the accuracy of the failure data (demand, run, and load, asappropriate) for the monitored components.
b. Findings
No findings of significance were identified.
Per TI 2515/169-05 reporting requirements, Attachment 2 to this report documentsadditional information pertaining to the inspectors review.
20Enclosure4OA6Meetings, Including ExitOn January 4, 2007, the inspectors presented the inspection results to Mr. Peter T.Dietrich and other members of his staff. The inspectors asked Entergy whether any ofthe material examined during the inspection should be considered proprietary. Noproprietary information was identified. 4OA7Licensee-identified ViolationsThe following violation of very low safety significance (Green) was identified by Entergyand is a violation of NRC requirements that met the criteria of Section VI of the NRCEnforcement Policy, NUREG-1600, for being dispositioned as an NCV.TS 5.4.1.a requires that procedures recommended in Appendix A of Regulatory Guide1.33 be implemented. Section 9.a of Regulatory Guide 1.33, Appendix A, requires thatmaintenance that can affect the performance of safety-related equipment be performedin accordance with written procedures appropriate to the circumstances. Entergyprocedure MP-023.14, "HPCI Turbine Minor Inspection, 23TU-2*," provides detailedinstructions for disconnecting and reconnecting control oil tubing from the HPCI turbineEG-R hydraulic actuator, including matchmarking of tubing connections andindependent verification. Contrary to this on October 25, 2006, two hydraulic lines to theEG-R actuator were connected to the wrong oil ports rendering the HPCI turbineinoperable. The condition was identified during post-work testing when the turbine failedsurveillance test ST-4N, "HPCI Quick Start, Inservice, and Transient Monitoring Test(IST)," and was documented in Entergy's corrective action program as CR-2006-04754. This finding is of very low safety significance because of the short duration of thecondition and the availability of all of the other redundant and diverse emergency corecooling systems. Documents reviewed for this inspection are listed in the Attachment 1.ATTACHMENT 1:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Entergy Personnel
- N. Avrakotos, Manager, Emergency Preparedness
- S. Bono, Director Engineering
- J. Costedio, Manager, Regulatory Compliance
- P. Dietrich, Site Vice President
- M. Durr, Manager, System Engineering
- M. Jacobs, Manager, Training
- D. Johnson, Manager, Operations
- J. LaPlante, Manager, Security
- K. Mulligan, General Manager, Plant Operations
- J. Pechacek, Manager, Programs and Components Engineering
- W. Rheaume, Manager, CA&A
- J. Solowski, Radiation Protection
- D. Wallace, Director, Nuclear Safety Assurance
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Closed
05000333/2002002-01URIAdequacy of Hemyc Cable Wrap FireBarrier Qualification Test and Evaluation(Section 4OA5.2)
LIST OF DOCUMENTS REVIEWED
Section 1R02: Evaluation of Changes, Tests, or ExperimentsSafety EvaluationsJAF-SE-00-003, "Update of
- FSAR to Remove Inconsistency Concerning Maximum EDG Room Temperature," Revision 0JAF-SE-00-025, "Provide Chemical Cleaning Process for ESW Piping and Heat Exchanger," Revision 4JAF-SE-01-009, "Feedwater Heaters Maximum String Flow," Revision 0JAF-SE-05-001, "ASME Code Repair of Containment (Torus)," Revision 0JAF-SE-05-002, "Evaluation of the Capability of Cooling Water Intake Bar Heaters," Revision 0