IR 05000382/2010005

Download: ML110420324

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Entergy, Inc. February 11, 2011 Joseph Kowalewski, Vice President, Operations Entergy Operations, Inc. Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-0751 Subject: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2010005

Dear Mr. Kowalewski:

On December 31, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed integrated inspection report documents the inspection findings, which were discussed on January 13, 2011, with you and other members of your staff.

The inspections 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. The report documents one self-revealing finding and one NRC-identified finding which were evaluated under the risk significance determination process as having very low safety significance (Green). These findings were determined to involve a violation of NRC requirements. Additionally, four licensee-identified violations which were also determined to be of very low safety significance are listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest any of the non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV; 612 East Lamar Blvd., Suite 400, Arlington, Texas 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspectors at the Waterford Steam Electric Station, Unit 3 facility. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent Entergy Operations, Inc. - 2 - possible, your response should not include any personal privacy or proprietary information so that it can be made available to the public without redaction.

Sincerely,/RA/ Jeffrey A. Clark, P.E. Chief, Project Branch E Division of Reactor Projects Docket Nos.: 50-382 License No.: NPF-38

Enclosure:

Inspection Report 05000382/2010005

w/Attachment:

Supplemental Information cc w/

Enclosure:

John T. Herron President and CEO Nuclear Operations/CNO Entergy Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Jeff Forbes Senior Vice President and Chief Operating Officer Entergy Operations, Inc. P. O. Box 31995 Jackson, MS 39286-1995 Thomas Palmisano Vice President, Oversight Entergy Operations, Inc. P. O. Box 31995 Jackson, MS 39286-1995 Senior Manager, Nuclear Safety and Licensing Entergy Services, Inc. P. O. Box 31995 Jackson, MS 39286-1995 Charles F. Arnone, General Manager, Plant Operations Entergy Operations, Inc. 17265 River Road Killona, LA 70057-0751 Steven Adams, Acting Director Nuclear Safety Assurance Entergy Operations, Inc. 17265 River Road Killona, LA 70057-0751 Billy Steeleman, Manager, Licensing and Regulatory Affairs Entergy Operations, Inc. 17265 River Road Killona, LA 70057-0751 Joseph A. Aluise Associate General Counsel - Nuclear Entergy Services, Inc. 639 Loyola Avenue New Orleans, LA 70113 Chairman Louisiana Public Service Commission P. O. Box 91154 Baton Rouge, LA 70821-9154 Parish President Council St. Charles Parish P. O. Box 302 Hahnville, LA 70057 St. Charles Parish Dept. of Emergency Preparedness Emergency Operations Center P.O. Box 302 Hahnville, LA 70057 Director, Nuclear Safety & Licensing Entergy, Operations, Inc. 440 Hamilton Avenue White Plains, NY 10601 Louisiana Department of Environmental Quality Radiological Emergency Planning and Response Division P. O. Box 4312 Baton Rouge, LA 70821-4312 Chief, Technological Hazards Branch FEMA Region VI 800 North Loop 288 Federal Regional Center Denton, TX 76209 Electronic distribution by RIV: Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Art.Howell@nrc.gov) DRP Director (Kriss.Kennedy@nrc.gov) DRP Deputy Director (Troy.Pruett@nrc.gov) DRS Director (Anton.Vegel@nrc.gov) Senior Resident Inspector (Marlone.Davis@nrc.gov) Resident Inspector (Dean.Overland@nrc.gov) Branch Chief, DRP/E (Jeff.Clark@nrc.gov) Senior Project Engineer, DRP/E (Ray.Azua@nrc.gov) Project Engineer (Jim.Melfi@nrc.gov) Project Engineer (Chris.Smith@nrc.gov) WAT Administrative Assistant (Linda.Dufrene@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Kaly.Kalyanam@nrc.gov) Branch Chief, DRS/TSB (Michael.Hay@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) Region IV RSLO (Bill.Maier@nrc.gov) NSIR/DPR/EP (Eric.Schrader@nrc.gov OEMail Resource Inspection Reports/MidCycle and EOC Letters to the following:

ROPreports Only inspection reports to the following: OEDO RIV Coordinator (James.Trapp@nrc.gov)

R:\_REACTORS\_WAT\2010\WAT 20100005 RPT.docx ADAMS ML110420324 ADAMS: No X Yes X SUNSI Review Complete Reviewer Initials: JAC X Publicly Available X Non-Sensitive Non-publicly Available Sensitive RIV:SRI:DRP/E RI:DRP/E SPE:DRP/E C:DRS/EB1 C:DRS/EB2 MFDavis DHOverland RVAzua TFarnholtz NFO'Keefe E-JClark E-JClark /RA/ /RA/ /RA/JMateychick for02/11/2011 02/11/2011 02/ 08/2011 02/ 08/2011 02/ 08/2011 C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C;DRS/TSB C:DRP/E MHaire MPShannon GEWerner MHay JAClark /RA/KClayton for /RA/ /RA/LCarson for/RA//RA/ 02/ /2011 02/ 08/2011 02/09/2011 02/09/2011 02/11/2011 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 50-382 License: NPF-38 Report: 05000382/2010005 Licensee: Entergy Operations, Inc. Facility: Waterford Steam Electric Station, Unit 3 Location: 17265 River Road Killona, LA 70057-0751 Dates: October 1, 2010 through December 31, 2010 Inspectors: M. Davis, Senior Resident Inspector D. Overland, Resident Inspector C. Steely, Operations Engineer L. Ricketson, Senior Health Physicist, PE N. Greene, PhD., Health Physicist P. Elkmann, Senior Emergency Preparedness Inspector R. Latta, Senior Reactor Inspector T. Burns, Senior Reactor Inspector, Region I D. Jones, Senior Reactor Inspector, Region III P. Prescott, Senior Quality and Vendor Programs Engineer, NRR Approved By: Jeffrey Clark, P.E. Chief, Project Branch E Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000382/2010005; 10/01/2010 - 12/31/2010; Waterford Steam Electric Station, Unit 3: Identification and Resolution of Problems. The report covered a three-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. The report also includes input from a 12 month NRC inspection of facilities for which Entergy Operations, Inc., holds a license, including the Waterford Steam Electric Station, Unit 3. One self-revealing finding and one NRC-identified finding, which were non-cited violations, were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process." The cross-cutting aspect was determined using Inspection Manual Chapter 0310, "Components Within the Cross Cutting Areas." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," occurred because the licensee did not promptly correct a significant condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of degraded diodes in static uninterruptible power supply A and B inverters, respectively. As a result, this led to another failure of a static uninterruptible power supply inverter. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0217. The immediate corrective actions following the additional failure included installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier. The planned corrective actions included implementation of an increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers. This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations" given the importance of the system and the fact that this condition affects static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its Technical Specification completion time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the decision-making component of human performance because the licensee did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. H.1(a) of IMC 0310] (Section 4OA2.4).

Green.

Inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion II, "Quality Assurance Program," for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that two individuals assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensee's overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the licensee's corrective action program as Condition Report CR-HQN-2010-00386. Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria." The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago. (Section 4OA2)

Other Findings

Licensee - Identified Violations Four violations of very low safety significance, which were identified by the licensee, have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and their associated corrective action tracking numbers are listed in Section 4OA7.

REPORT DETAILS

Summary of Plant Status The Waterford Steam Electric Station, Unit 3, began the inspection period at 100 percent power and remained at approximately 100 percent power for the rest of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather Protection

Readiness to Cope with External Flooding

a. Inspection Scope

The inspectors evaluated the design, material condition, and procedures for coping with the design basis probable maximum flood. The evaluation included a review to check for deviations from the descriptions provided in the updated final safety analysis report for features intended to mitigate the potential for flooding from external factors. As part of this evaluation, the inspectors checked for obstructions that could prevent draining, checked that the roofs did not contain obvious loose items that could clog drains in the event of heavy precipitation, and determined that barriers required to mitigate the flood were in place and operable. Additionally, the inspectors performed an inspection of the protected area to identify any modification to the site that would inhibit site drainage during a probable maximum precipitation event or allow water ingress past a barrier. The inspectors also reviewed the abnormal operating procedure for mitigating the design basis flood to ensure it could be implemented as written. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one (1) external flooding sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignments

Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• On October 12, 2010, Train B of the essential chiller during a scheduled maintenance outage of Train A

• On November 19, 2010, Train A of the auxiliary component cooling water system during a scheduled maintenance outage of Train B The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, updated final safety analysis report, technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also inspected accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of two (2) partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• On October 1, 2010, reactor auxiliary building (RAB) wing area, fire area RAB 32

• On October 20, 2010, Train B emergency diesel generator area, fire area RAB 15

• On November 17, 2010, Train B vital switchgear area, fire area RAB 8b

• On December 12, 2010, turbine generator building +15 elevation The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensee's fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plant's ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensee's corrective action program. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four (4) quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings were identified.

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On November 10, 2010, the inspectors observed the fire brigade activation in the +15 turbine generator building west fire area. The observation evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were (1) proper wearing of turnout gear and self-contained breathing apparatus; (2) proper use and layout of fire hoses; (3) employment of appropriate fire fighting techniques; (4) sufficient firefighting equipment brought to the scene; (5) effectiveness of fire brigade leader communications, command, and control; (6) search for victims and propagation of the fire into other plant areas; (7) smoke removal operations; (8) utilization of preplanned strategies; (9) adherence to the preplanned drill scenario; and (10) drill objectives.

These activities constitute completion of one (1) annual fire-protection inspection sample as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

The inspectors reviewed the updated final safety analysis report, the flooding analysis, and plant procedures to assess susceptibilities involving internal flooding; reviewed the corrective action program to determine if licensee personnel identified and corrected flooding problems; inspected underground bunkers/manholes to verify the adequacy of sump pumps, level alarm circuits, cable splices subject to submergence, and drainage for bunkers/manholes; and verified that operator actions for coping with flooding can reasonably achieve the desired outcomes. The inspectors also inspected the areas listed below to verify the adequacy of equipment seals located below the flood line, floor and wall penetration seals, watertight door seals, common drain lines and sumps, sump pumps, level alarms, and control circuits, and temporary or removable flood barriers. Specific documents reviewed during this inspection are listed in the attachment.

• On October 28, 2010, turbine cooling water underground cable inspections

• On December 14, 2010, component cooling water pump rooms These activities constitute completion of two (2) flood protection measures inspection samples as defined in Inspection Procedure 71111.06-05.

b. Findings

Introduction.

The inspectors identified an issue of concern related to debris entering the condensate storage pool from an eight-inch diameter vent line located in the component cooling water pump room B.

Description.

On December 16, 2010, during inspections of the component cooling water pump rooms for flooding protection, the inspectors identified several concerns. The inspectors noticed an 8 inch diameter, 6 foot tall gooseneck vent line in component cooling water pump room B that was not included or discussed in the flooding evaluation provided in Updated Final Safety Analysis Report section 3.6A.6.4.2.3. The inspectors were concerned for debris possibly entering the condensate storage pool from component cooling water pump Room B. The inspectors did not notice any screen, design feature, or foreign material exclusion device on the vent. Updated Final Safety Analysis Report section 10.4.9B.3.2 discusses common cause failures for the emergency feedwater system, whose pumps take suction on the condensate storage pool. Updated Final Safety Analysis Report section 10.4.9B.3.2.g states, in part, that particles/corrosion associated with condensate storage pool would be small and not credible. The vent line is an 8 inch open pathway into the condensate storage pool. The inspectors have not reviewed the licensee's foreign material exclusion procedure(s), or note what the maximum particle size that is allowed in the condensate storage pool.

This issue of concern will be treated as an unresolved item pending the review of additional information from the licensee related to a potential for a common mode failure of the emergency feedwater system due to debris: URI 05000382/2010005-01; Foreign Material Exclusion Issues associated with the Condensate Storage Pool Gooseneck Vent.

1R07 Heat Sink Performance

a. Inspection Scope

The inspectors reviewed licensee programs, verified performance against industry standards, and reviewed critical operating parameters and maintenance records for the Train B shutdown cooling heat exchanger. The inspectors verified that performance tests were satisfactorily conducted for heat exchangers/heat sinks and reviewed for problems or errors; the licensee utilized the periodic maintenance method outlined in EPRI Report NP 7552, "Heat Exchanger Performance Monitoring Guidelines;" the licensee properly utilized biofouling controls; the licensee's heat exchanger inspections adequately assessed the state of cleanliness of their tubes; and the heat exchanger was correctly categorized under 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one (1) heat sink inspection sample as defined in Inspection Procedure 71111.07-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Quarterly Review

a. Inspection Scope

On November 29, 2010, the inspectors observed a crew of licensed operators in the plant's simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crew's clarity and formality of communications

• Crew's ability to take timely actions in the conservative direction

• Crew's prioritization, interpretation, and verification of annunciator alarms

• Crew's correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crew's ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications The inspectors compared the crew's performance in these areas to preestablished operator action expectations and successful critical task completion requirements. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one (1) quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Annual Inspection

a. Inspection Scope

The inspectors reviewed the annual operating test results for 2010. Since this was the first half of the biennial requalification cycle, the licensee was not required to administer a written examination. These results were assessed to determine if they were consistent with NUREG 1021, "Operator Licensing Examination Standards for Power Reactors," guidance and Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," thresholds. This review included the test results for a total of 10 crews composed of 39 senior reactor operators and 12 reactor operators. All individuals and crews passed all portions of the operating test. The inspectors completed one (1) inspection sample of the biennial licensed operator requalification program.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk significant systems:

• On November 22, 2010, control room air handling units relay cards

• On December 10, 2010, dry cooling tower fans The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1) The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of two (2) quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings were identified.

R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed licensee personnel's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• On November 10, 2010, emergent maintenance on the 'A' static uninterruptable power supply inverter with the Train A emergency boration makeup pump and essential feedwater valve (223B) out of service. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one (1) maintenance risk assessments and emergent work control inspection sample as defined in Inspection Procedure 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• On October 12, 2010, operability evaluation related to the air leak on atmospheric dump Valve, MS-116B.

• On October 22, 2010, operability evaluation related to elevated wattage on Train B startup transformer center phase

• On December 13, 2010, operability evaluation related to 3 inch void discovery in safety injection Train A discharge line The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and updated final safety analysis report to the licensee personnel's evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of three (3) operability evaluations inspection sample(s) as defined in Inspection Procedure 71111.15-04

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Modifications

a. Inspection Scope

To verify that the safety functions of important safety systems were not degraded, The inspectors reviewed the following temporary modifications:

• On December 13, 2010, temporary modification to install a refueling machine jumper

• On December 13, 2010, temporary modification to install a time delay in the open circuit for Valves SI-401A and SI-401B The inspectors reviewed the temporary modifications and the associated safety-evaluation screening against the system design bases documentation, including the updated final safety analysts report and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors verified that the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers. These activities constitute completion of two (2) samples for temporary plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• On September 30, 2010, replaced a faulty relay on the control room air handling unit 12B

• On October 8, 2010, scheduled outage for the A chemical volume control pump

• On November 5, 2010, scheduled outage of the A emergency diesel generator

• On November 10, 2010, replaced diodes and silicon control rectifier on the 'A' static uninterruptable power supply inverter

• On December 13, 2010, scheduled outage of the AB high pressure safety injection to calibrate thermal overload relays for safety injection valves SI-226B, SI-227B, and SI-502B The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following (as applicable):

• The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate The inspectors evaluated the activities against the technical specifications, the updated final safety analysis report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five (5) postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, procedure requirements, and technical specifications to ensure that the surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• On October 7, 2010, scheduled surveillance to verify operability of the chilled water pump AB

• On November 19, 2010, scheduled surveillance to verify operability of auxiliary component cooling water pump B

• On December 13, 2010, containment isolation valves SI-226B, SI-227B, and SI-502B demonstrated operable prior to returning valves to service Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three (3) surveillance testing inspection samples including one containment isolation valve sample as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstone:

Occupational and Public Radiation Safety

2RS0 1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

This area was inspected to: (1) review and assess licensee's performance in assessing the radiological hazards in the workplace associated with licensed activities and the implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures, (2) verify the licensee is properly identifying and reporting Occupational Radiation Safety Cornerstone performance indicators, and (3) identify those performance deficiencies that were reportable as a performance indicator and which may have represented a substantial potential for overexposure of the worker.

The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensee's procedures required by technical specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors performed walkdowns of various portions of the plant, performed independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation reported by the licensee in the Occupational Radiation Safety Cornerstone

• The hazard assessment program, including a review of the license's evaluations of changes in plant operations and radiological surveys to detect dose rates, airborne radioactivity, and surface contamination levels

• Instructions and notices to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions

• Programs and processes for control of sealed sources and release of potentially contaminated material from the radiologically controlled area, including survey performance, instrument sensitivity, release criteria, procedural guidance, and sealed source accountability

• Radiological hazards control and work coverage, including the adequacy of surveys, radiation protection job coverage, and contamination controls; the use of electronic dosimeters in high noise areas; dosimetry placement; airborne radioactivity monitoring; controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools; and posting and physical controls for high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements

• Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of the one (1) required sample as defined in Inspection Procedure 71124.01-05.

b. Findings

No findings were identified.

2RS0 2 Occupational ALARA Planning and Controls

a. Inspection Scope

This area was inspected to assess performance with respect to maintaining occupational individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensee's procedures required by technical specifications as criteria for determining compliance. During the inspection, the inspectors interviewed licensee personnel and reviewed the following items:

• Site-specific ALARA procedures and collective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements

• ALARA work activity evaluations/post job reviews, exposure estimates, and exposure mitigation requirements

• The methodology for estimating work activity exposures, the intended dose outcome, the accuracy of dose rate and man-hour estimates, and intended versus actual work activity doses and the reasons for any inconsistencies

• Records detailing the historical trends and current status of tracked plant source terms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

• Audits, self-assessments, and corrective action documents related to ALARA planning and controls since the last inspection Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of the one required sample as defined in Inspection Procedure 71124.02-05.

b. Findings

Introduction.

The inspectors identified an issue of concern related to the control of occupational radiation exposures during the most recent refueling outage. This issue is an unresolved item pending a review of additional information from the licensee.

Description.

On November 15, 2010, the inspectors reviewed 13 radiation work permit packages for work activities performed during Refueling Outage 16. Of the 13 radiation work permit packages reviewed, 12 had an actual dose value of greater than 5 rem and exceeded their original dose estimate by more than 50 percent. Two of the 12 radiation work permit packages had an actual dose value greater than 25 rem. The inspectors interviewed licensee personnel and reviewed post job reviews to determine the reason the collective dose exceeded the dose estimates. The higher doses were primarily caused by a hard trip (i.e., thermal hydraulic shock) from 100 percent power resulting in CRUD burst and plate out in the reactor coolant system and reactor coolant pump seal leakage due to destaging of the reactor coolant pump seal vapor stages when the reactor coolant system pressure fell below 350 pounds.

In the original outage plan, the reactor was scheduled to gradually reduce power to 80 percent three days prior to shutdown. However, as a result of a moisture separator reheater relief valve failing to open, the reactor was tripped from 100 percent power on October 19, 2009. The licensee performed a root cause evaluation and determined the problem resulted from a cracked and dislocated compressed spring. The crack was caused by a problem during the manufacturing process. The inspectors concluded the licensee could not have foreseen and prevented this problem. However, according to the apparent cause evaluation associated with Condition Report CR-WF3-2009-07262, the steam generators were not drained in the usual timeframe as in all previous outages. In previous outages, the steam generators were usually drained by the fourth day of the outage. During Refueling Outage 16, the steam generators were not drained until day 17. Shutdown cooling was initiated on October 20, 2009, and resulted in low flow of water through the steam generators and little clean up. This combined with a Cobalt-58 peak of 5.0 microcuries per milliliters allowed settling and further plate out of CRUD in the steam generators. This was an issue of concern because the actual radioactive concentration of Cobalt-58 was considerably greater than the recommended target value of less than 0.05 microcuries per milliliter for Cobalt-58 as listed in Procedure CE 006, "Maintaining Reactor Coolant System Chemistry," Revision 305, and could have contributed to the higher than anticipated work activity doses. The inspectors concluded more information was required related to the licensee's decision to not remove additional radioactivity from the steam generators before the inspectors could determine if a performance deficiency existed. The inspectors asked the licensee to provide the required information during a teleconference conducted December 1, 2010. This issue of concern will be treated as an unresolved item pending the review by the inspectors of additional information from the licensee related to the licensee's decision to not remove additional radioactivity from the steam generators: URI 05000382/2010005-02; Removal of Radioactivity from the Steam Generators. On October 23, 2009, during RCS depressurization to reach mode 5, all four reactor coolant pumps seal vapor stages de-staged at approximately 350 psia and leaked highly contaminated reactor coolant system water onto pump insulation, adjacent structures, and onto the -11 foot elevation of the reactor containment building. As a result of this leakage, dose rates in the immediate vicinity of the reactor coolant pumps were elevated to 3.5 rems per hour. Contamination levels were as high as 500 millirads per hour. This leakage was anticipated by the licensee because, according to CR-WF3-2007-03716, the phenomenon was first observed September 2005. The inspectors identified the licensee's failure to prevent the leakage of radioactively contaminated water from the reactor coolant pumps as an issue of concern which could have contributed to the higher than anticipated work activity doses, but concluded additional information regarding the design of the reactor coolant pump seals and licensee's opportunity to correct the problem of leakage from the seals was required to determine if a performance deficiency existed. The inspectors asked the licensee to provide the required information during a teleconference conducted December 1, 2010. This issue of concern will be treated as an unresolved item pending the review by the inspectors of additional information from the licensee related to the design of the reactor coolant pump seals and licensee's opportunity to correct the problem of leakage from the seals: URI 05000382/2010005-03; Leakage from the Reactor Coolant Pump Seals.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the performance indicator data submitted by the licensee for the fourth Quarter 2010 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, "Performance Indicator Program." This review was performed as part of the inspectors' normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings were identified.

.2 Safety System Functional Failures (MS05)

a. Inspection Scope

The inspectors sampled licensee submittals for the safety system functional failures performance indicator for the period from the third quarter 2009 through the third quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, and NUREG-1022, "Event Reporting Guidelines 10 CFR 50.72 and 50.73." The inspectors reviewed the licensee's operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports, and NRC integrated inspection reports for the period of March 2009 through September 2010 to validate the accuracy of the submittals. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one (1) safety system functional failures sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.3 Mitigating Systems Performance Index - Emergency ac Power System (MS06)

a. Inspection Scope

The inspectors sampled licensee submittals for the mitigating systems performance index - emergency ac power system performance indicator for the period from the third quarter 2009 through the third quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6. The inspectors reviewed the licensee's operator narrative logs, mitigating systems performance index derivation reports, issue reports, event reports, and NRC integrated inspection reports for the period of third quarter 2009 through the third quarter 2010 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one (1) mitigating systems performance index emergency ac power system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.4 Mitigating Systems Performance Index - Cooling Water Systems (MS10)

a. Inspection Scope

The inspectors sampled licensee submittals for the mitigating systems performance index - cooling water systems performance indicator for the period from the third quarter 2009 through the third quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6. The inspectors reviewed the licensee's operator narrative logs, issue reports, mitigating systems performance index derivation reports, event reports, and NRC integrated inspection reports for the period of third quarter 2009 through the third quarter 2010 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance.

The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one (1) mitigating systems performance index cooling water system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.5 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

Cornerstone: Occupational Radiation Safety The inspectors reviewed performance indicator data for the third quarter 2009 through the third quarter 2010. The objective of the inspection was to determine the accuracy and completeness of the performance indicator data reported during these periods. The inspectors used the definitions and clarifying notes contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, as criteria for determining whether the licensee was in compliance.

The inspectors reviewed corrective action program records associated with high radiation area (greater than 1 rem/hr) and very high radiation area nonconformances. The inspectors reviewed radiological, controlled area exit transactions greater than 100 mrem. The inspectors also conducted walkdowns of high radiation areas (greater than 1 rem/hr) and very high radiation area entrances to determine the adequacy of the controls of these areas. These activities constitute completion of the one (1) occupational exposure control effectiveness sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.6 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (PR01)

a. Inspection Scope

Cornerstone: Public Radiation Safety The inspectors reviewed performance indicator data for the third quarter 2009 through the third quarter 2010. The objective of the inspection was to determine the accuracy and completeness of the performance indicator data reported during these periods. The inspectors used the definitions and clarifying notes contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, as criteria for determining whether the licensee was in compliance.

The inspectors reviewed the licensee's corrective action program records and selected individual annual or special reports to identify potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose.

These activities constitute completion of the one (1) radiological effluent technical specifications/offsite dose calculation manual radiological effluent occurrences sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensee's corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensee's corrective action program because of the inspectors' observations are included in the attached list of documents reviewed.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure, they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's corrective action program. The inspectors accomplished this through review of the station's daily corrective action documents. The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensee's corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human performance results. The inspectors nominally considered the 6-month period of July 2010 through December 2010 although some examples expanded beyond those dates where the scope of the trend warranted. The inspectors also included issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments. The inspectors compared and contrasted their results with the results contained in the licensee's corrective action program trending reports. Corrective actions associated with a sample of the issues identified in the licensee's trending reports were reviewed for adequacy. These activities constitute completion of one (1) single semi-annual trend inspection sample as defined in Inspection Procedure 71152-05.

b. Findings

No findings were identified.

.4 Selected Issue Follow-up Inspection: Static Uninterruptable Power Supply Inverters

a. Inspection Scope

The inspectors performed an in-depth review of the licensee's evaluation and corrective actions related to additional failures of the static uninterruptable power supply inverters A and B, respectively. The inspectors reviewed the appropriateness of the assigned significance, the scope and depth of the causal analysis, and the timeliness of resolution. The inspectors assessed whether the evaluation identified likely causes for the issues and identified appropriate corrective actions to address the identified causes. The inspectors also conducted a review of the corrective actions to verify that appropriate measures were in place to prevent reoccurrence of the issue. In addition, the inspectors assessed whether the licensee's evaluation considered extent of condition, generic implications, common cause, and previous occurrences. The inspectors reviewed the potential impact on nuclear safety and risk to verify that the licensee had taken corrective actions commensurate with the significance of the issue. The inspectors evaluated these actions against the requirements of the licensee's corrective actions program and performance attributes contained in IP 71152, Section 03.06. These activities constitute completion of one (1) in-depth problem identification and resolution sample as defined in Inspection Procedure 71152-05.

b. Findings

Introduction.

A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," occurred because the licensee did not promptly correct a significant condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of potentially degraded diodes in static uninterruptible power supply A and B inverters, respectively. As a result, this led to another failure of the static uninterruptible power supply A inverter.

Description.

On February 14, 2009, the Waterford Steam Electric Station, Unit 3 experienced a static uninterruptible power supply A inverter failure that caused the plant to enter an unplanned 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown limiting condition of operation. The licensee initiated a higher tier apparent cause evaluation CR-WF3-2009-0802 and determined that the apparent cause was a short in one of the inverter's bridge power switching diodes. The immediate corrective action was to replace the diode. The planned corrective actions were to evaluate the failure mode of the diode and revise the preventive maintenance program for static uninterruptible power supply inverters, as necessary, to ensure it was adequate. These actions were ultimately not effective in preventing a second failure because on April 11, 2010, the static uninterruptible power supply B inverter failed. The licensee determined that the B inverter failure was related to a short in one of the inverter's bridge power switching diodes, which was similar to the A inverter failure. The licensee documented this issue as a significant condition adverse to quality in CR-WF3-2010-2278 and conducted a root cause evaluation. The licensee concluded that the manufacturer fabrication and construction of the bridge power switching diodes were less than adequate and that the preventive maintenance and component monitoring were contributing causes. The licensee developed a corrective action plan to preclude repetition of this significant condition adverse to quality, which was to install new diodes and silicon controlled rectifiers with a new manufacturer and perform increase component monitoring. In addition, the licensee added requirements to their receipt inspection to perform pre-installation testing on spare diodes and silicon controlled rectifiers prior to installing new diodes in safety-related inverters. The licensee decided to install these new diodes and silicon controlled rectifiers with a new manufacturer in refueling outage 17, scheduled for earlier spring of 2011. However, on November 10, 2010, the plant experienced another failure of the static uninterruptible power supply A inverter. The cause was similar to the two previous failure mechanisms, which was a short in one of the inverter's bridge power switching diodes. The inspectors reviewed the root cause evaluation report (CR-WF3-2010-2278, Revision 1) from July of 2010, associated condition reports, work orders, and other related documents. The inspectors also interviewed site personnel associated with this issue and the justification to extend the corrective actions to refueling outage 17. The justification to perform this as a long term corrective action stated that this was deemed an outage task. However, the inspectors noted that there was no other justification as to why this was deemed an outage task or what was necessary to implement the corrective action. The inspectors noted that there was no discussion on the safety significance and the significance of the degradation in extending the corrective action to a refueling outage. The inspectors determined that the licensee did not appropriately justify a longer completion schedule to perform corrective actions on these safety-related inverters. In fact, a review of the licensee's long term corrective action plan, stated in part, that maintenance personnel prepared a model work order to allow for quick troubleshooting of a diode failure and had additional spare diodes and silicon controlled rectifiers in stock such that if there was a diode failure then it could have been corrected within its technical specification limiting condition of operation completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee performed immediate corrective actions following the additional failure that included the installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier within the allowed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> technical specification completion time. The planned corrective actions included implementation of an increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers.

Analysis.

The performance deficiency is that the licensee did not promptly correct a significant condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not replace degraded diodes on safety-related inverters in a timely manner because the licensee did not appropriately justify a longer completion schedule to perform these corrective actions. As a result, this led to an additional static uninterruptible power supply inverter failure. This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations" given the importance of the system and the fact that this condition affects static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its Technical Specification completion time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the decision-making component of human performance because the licensee did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. H.1(a) of IMC 0310].

Enforcement.

Title 10 of CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, prior to November 10, 2010, the licensee did not promptly correct a significant condition adverse to quality. Specifically, the licensee did not appropriately justify a longer completion schedule to perform corrective actions on the static uninterruptible power supply inverters. As a result, this led to an additional static uninterruptible power supply inverter failure. However, because this finding was of very low safety significance and it was entered into the corrective action program as CR-WF3-2011-0217, this violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000382/2010005-04: Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters)

.5 Selected Issue Follow-up Inspection: Review of Circumstances Surrounding Missed Quality Control Verification Inspections

a. Inspection Scope

An inspection was performed at the Entergy corporate office in Jackson, Mississippi on June 14 through 17, 2010, to review the circumstances surrounding missed quality control (QC) verification inspections documented in CR-HQN-2009-01184 and CR-HQN-2010-00013. The issue involved QC verification inspections performed during construction-related activities which were required as part of the Entergy quality oversight and verification programs. The inspection was performed to determine if the licensee had taken corrective actions commensurate with the significance of the identified issues, and to assess the impact, if any, on the operability of plant equipment caused by the missed inspections. This inspection was conducted by inspectors from Regions I, II, and IV, as well as a Senior Program Engineer from the Quality and Vendor Branch of the Office of Nuclear Reactor Regulation (NRR). The inspection covered all NRC-licensed sites owned by Entergy Operations, Inc., including Arkansas Nuclear One, James A. Fitzpatrick, Grand Gulf Nuclear Station, Indian Point Units 2 and 3, Palisades Plant, Pilgrim Nuclear Power Station, River Bend Station, Vermont Yankee, and Waterford Steam Electric Station, Unit 3.

The inspectors reviewed root cause analyses documented in Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-00013, and the results of the licensee's extent of condition reviews and plant impact assessments. The inspectors also independently assessed the potential impacts of the missed inspections on the operability of plant equipment by reviewing all of the examples identified by the licensee, and by independently reviewing completed modifications and work orders to identify additional examples. The inspectors also reviewed the corrective action database to assess reported equipment failures in order to assess whether the failure might have involved missed QC verification inspections. The inspectors assessed causal factors that may have contributed to missing QC verification inspections. This assessment included reviewing the Entergy Quality Assurance Program Manual (QAPM) requirements, changes made to the QAPM, and the level of agreement between the QAPM and its implementing procedures. Specific documents reviewed are listed in the attachment.

b. Findings

Background The inspectors identified problems with the implementation of elements of the Quality Assurance (QA) Program that affected the fleet of Entergy Operations Inc., (hereafter referred to as "Entergy") nuclear power plants that are licensed by the NRC. While the plant organizations are NRC licensees, Entergy also has corporate groups which are not NRC licensees that are actively involved in some activities affecting sites, including program and procedure changes. Entergy adopted a business strategy of adopting standard programs and procedures at all fleet plants.

On October 30, 2009, the NRC discussed with Entergy the initial concerns about whether QC verification inspections were being performed consistently for the types of work that require that level of inspection. Both the non-licensed and licensed Entergy organizations responded with an appropriate review of the issues. Entergy's review of work documents that were potentially affected was extensive at each site. Entergy's total review examined over 320 Engineering Change documents and 2676 Work Orders. Of the 30 Work Orders identified to have QC verification inspection deficiencies affecting eight safety-related design changes, all 30 were determined by Entergy to have sufficient documentation to provide confidence that the equipment was installed correctly. Specific corrective actions were identified and implemented to ensure that QC verification inspections would be included in current and future work documents, including procedure enhancements.

The information provided to the NRC was used to perform a focused inspection in order to assess the impact of the missed verification inspections at each of the NRC-licensed facilities. The inspection documented below independently assessed the potential impact of missed QC verification inspections on the operability of plant equipment, as well as assessing details of the QA Program for the Entergy fleet.

Two findings were identified during this inspection. These findings involved missed QC verification inspections at seven Entergy sites, and the assignment of individuals to the QA Manager position that did not meet the experience and qualification requirements at eight sites. Only the findings impacting this licensee are described below.

The inspectors concluded that the Entergy fleet organizational structure and Entergy strategy of adopting standardized procedures across the fleet were contributing factors to the findings. Specifically:

• Changes to adopt the standard fleet QA program created a partially conflict with existing requirements for worker qualifications at some sites. The process for creating and revising standardized fleet procedures and programs used to meet NRC requirements must ensure that site-specific regulatory requirements and commitments are properly addressed for all sites.

• Changes that removed details from existing site-specific QA and QC program implementing procedures while shifting to standardized fleet procedures contributed to the finding involving missed QC verification inspections. Condition reports at individual sites regarding problems related to this issue were not recognized collectively as symptoms of a problem with these procedures because they were addressed at the site level.

Failure to Implement the Experience and Qualification Requirements Associated With the Quality Assurance Program

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion II, "Quality Assurance Program," for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that two individuals assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.

Description.

During their review of the issues surrounding the improper implementation of quality control (QC) verifications discussed above, the inspectors noted that the root cause analysis documented in CR-HQN-2010-0013 identified that lack of experience of the Quality Assurance (QA) Manager contributed to the failure to identify the trend in missed QC verification inspections. The inspectors reviewed the relevant experience and qualifications of the QA Manager at each Entergy site. The inspectors also reviewed the NRC's safety evaluation report that approved Entergy's original corporate Quality Assurance Program Manual (QAPM), which is the document that contains the QA Program. Additionally, the inspectors reviewed the administrative section of the Technical Specifications for all the Entergy sites and a sample of evaluations, performed in accordance with 10 CFR 50.54(a), that supported Entergy QAPM changes and alignment of plants that were subsequently purchased by Entergy.

The Entergy corporate QAPM required each site to meet the experience and qualification standards in ANSI/ANS 3.1-1978, "American National Standard for Selection and Training of Nuclear Power Plant Personnel." Section 4.4 included qualification and experience requirements for the personnel described as "group leaders" of five professional-technical groups, including Quality Assurance. Section 4.4.5, "Quality Assurance," required that "-the responsible person shall have six years experience in the field of quality assurance, preferably at an operating nuclear plant, or operations supervisory experience. At least one year of this six years experience shall be nuclear power plant experience in the overall implementation of the quality assurance program. (This experience shall be obtained within the quality assurance organization.)" On December 15, 2008, procedure EN-QV-117, "Oversight Training Program," the Entergy procedure used by all Entergy sites to implement the requirements of ANSI/ANS 3.1-1978, was revised by the Entergy corporate QA group. Section 5.7, "Manager/QA Senior Auditor Training," was changed to state: Either the QA Manager or the Senior QA Auditor will meet the requirements of ANS 3.1-1978 paragraph 4.4.5 for operating plants and if applicable ANS 3.1-1993 paragraph 4.3.7 for new plants. The inspectors reviewed completed Personnel Change Planning Checklist/Forms for QA Managers at each site. Entergy used this form to evaluate QA Manager candidates prior to the implementation of an Entergy fleet-wide restructuring in July 2007. Attachment 8, "Change Management Guidelines for Alignment Implementation," included the following conclusion for the individual that subsequently was assigned to be the QA Manager: [Individual's name redacted] meets the minimum requirements for QA Manager with the exception of at least one year of this six years experience shall be nuclear power plant experience in the overall implementation of the quality assurance program. This requirement must be met by the QA Senior Auditor. Based on discussions with Entergy corporate QA personnel, the inspectors determined that Entergy personnel had interpreted ANSI/ANS 3.1-1978, Sections 4.4 and 4.4.5 to allow the Senior Auditor to be considered the QA group leader described in the standard for purposes of meeting the experience requirements of Section 4.4.5 in cases where a candidate for the position of QA Manager did not satisfy the experience requirements.

In reviewing this issue, the NRC staff has determined that the group leader in this case is the individual filling the position assigned responsibility for overall implementation of the QA Program (Entergy used the title "QA Manager" for this position). The individual meeting the experience and qualification requirements must be the individual assigned the responsibilities for overall implementation of the QA Program assigned within the QA Program.

The inspectors determined that this change to procedure EN-QV-117 did not ensure that the qualifications for the QA Manager would meet the requirements of standard. The inspectors identified two examples where the Senior Auditor was credited as being the group leader for purposes of meeting ANSI/ANS 3.1-1978, and the individuals who were assigned as the QA Manager did not meet the ANSI/ANS 3.1-1978 experience requirements. The team also determined that the responsibilities assigned to the QA Manager under the QAPM were not reassigned to the Senior Auditor, and the Senior Auditor did not report directly to the designated senior executive. The Senior Auditor continued to report to the QA Manager, so the person with the greater experience did not have the positional authority to decide issues.

Analysis.

Failure to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the QA Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, so it was evaluated using the qualitative criteria listed in Table 4.1. This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist.

The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.

Enforcement.

Appendix B to 10 CFR 50, Criterion II, "Quality Assurance Program," requires, in part, that the licensee establish a quality assurance program which complies with Appendix B. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions. The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained. The Entergy Quality Assurance Program Manual, Revision 13, is the document used at each Entergy-owned site to describe the quality assurance program. Table 1, Section A of the Quality Assurance Program Manual states, in part, that qualifications and experience for station personnel shall meet ANSI/ANS 3.1-1978 except for positions where an exception to either ANSI/ANS 3.1-1978 or N18.1-1971 is stated in the applicable unit's Technical Specifications. ANSI/ANS 3.1-1978, Section 4.4.5, "Quality Assurance," states, in part, that the responsible person (i.e. the Quality Assurance Manager) shall have six years experience in the field of quality assurance. At least one year of this six years experience shall be obtained within the quality assurance organization.

Contrary to the above, between May 15, 2005, and May 15, 2006, and between July 7, 2007 and July 7, 2008, the licensee failed to implement the quality assurance program requirements intended to provide indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency was achieved and maintained. Specifically, the individual(s) assigned to be the responsible person for the licensee's overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. Because this issue was of very low safety significance and was entered into the corrective action program as Condition Report CR-HQN-2010-00386, consistent with Section 2.3.2.a of the Enforcement Policy, this violation is being treated as a non-cited violation, NCV 05000382/2010005-05: Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program.

4OA3 Event Follow-up

.1 (Closed) Licensee Event Report (LER) 05000382/2009001-00, Mode Change with an Inoperable Emergency Feedwater Pump On April 27, 2009, the licensee identified that during a startup in May 2008, the plant entered Mode 3 with emergency feedwater Pump AB inoperable due to a lifted lead on

the governor control system. The mode change was not in conformance with technical specification 3.0.4, which precludes a mode shift with the emergency feedwater pump inoperable. The licensee also failed to recognize the reportable condition until almost a year later. The inspectors determined that these conditions occurred as a result of three examples of a failure to maintain and follow safety-related procedures. First, the work instruction governing maintenance on the pump contained incorrect guidance. Second, the step in EN-OP-104 to verify operability of all equipment required for the mode shift was signed as complete, when all required equipment was not operable. Third, the licensee failed to follow EN-LI-102 and properly recognize the reportability of the event. This licensee-identified finding involved a violation of technical specification 6.8.1.a, which requires the licensee to maintain and implement procedures recommended in Regulatory Guide 1.33. The safety significance and enforcement aspects of the violation are discussed in Section 4OA7. This licensee event report is closed.

.2 (Closed) Licensee Event Report (LER) 05000382/2009004-00, Condition Prohibited by Technical Specification with Log Power Channel Inoperable On July 5, 2009, the inspector identified a non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation.

The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. This violation was documented as05000382/2009004-1. This licensee event report is closed.

OA5 Other Activities

.1 (Closed) Temporary Instructions 2515/180, Inspection of Procedures and Processes for Managing Fatigue

a. Inspection Scope

The inspectors reviewed the licensee's procedures and policies to confirm that the Fitness for Duty program adequately implemented fatigue management requirements for individuals subject to 10 CRF 26, Subpart I. The inspectors confirmed that the licensee had procedures in place that described:

• The process to be followed after any individual makes a self-declaration that he or she is not fit to safely and competently perform his or her duties for any part of a working tour as a result of fatigue;

• The process for implementing the work hour controls;

• The process for conducting fatigue assessments, and

• Disciplinary actions that may be imposed on an individual following a fatigue assessment, and the conditions and considerations for taking those disciplinary actions. The inspectors reviewed the licensee's training program to verify the implementation and testing of specified knowledge and abilities specified in 10 CFR 26.203(c)(1) and (c)(2). The inspectors confirmed that the licensees' process for developing the annual Fitness for Duty report include provisions for documenting the summary of instances where work hour controls were waived. The inspectors also confirmed that the licensee had a process in place to retain the following records for at least 3 years or until the completion of all related legal proceedings, whichever is later:

• Work hours for individuals who are subject to the work hour controls;

• Shift schedules and shift cycles of individuals who are subject to the work hour controls;

• Waivers and the bases for the waivers,

• Work hour reviews; and

• Fatigue assessments. These activities constitute completion of Temporary Instruction 2515/180, Inspection of Procedures and Processes for Managing Fatigue.

b. Findings

Overall, the licensee implementation of the fatigue management requirements was adequate. However, the inspectors identified issues of concern in the licensee's security department area. The NRC security inspection report 05000382/2010403 documents the results of these deficiencies.

.2 (Closed) Temporary Instruction (TI) 2515/179, "Verification of Licensee Responses to NRC Requirement for Inventories of Materials Tracked in the National Source Tracking System Pursuant to Title 10, Code of Federal Regulations, Part 20.2207 (10 CFR 20.2207)"

a. Inspection Scope

An NRC inspection was performed to confirm that the licensee has reported their initial inventories of sealed sources pursuant to 10 CFR 20.2207 and to verify that the National Source Tracking System database correctly reflects the Category 1 and 2 sealed sources in custody of the licensee. Inspectors interviewed personnel and performed the following:

• Reviewed the licensee's source inventory

• Verified the presence of any Category 1 or 2 sources

• Reviewed procedures for and evaluated the effectiveness of storage and handling of sources

• Reviewed documents involving transactions of sources

• Reviewed adequacy of licensee maintenance, posting, and labeling of nationally tracked sources

b. Findings

No findings were identified

4OA6 Meetings Exit Meeting Summary

On October 27, 2010, the inspectors discussed the inspection results of the licensed operator requalification program annual operating test with Mr. J. Signorelli, Operations Training Instructor. The licensee acknowledged the results. The inspectors confirmed that proprietary information was not provided or examined during the inspection. On November 10, 2010, the inspectors conducted a telephonic exit meeting to present the results of in-office inspection of the licensee's implementation of emergency action levels to Mr.

R. Murillo, Acting Director, Nuclear Safety Assurance, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. On November 18, 2010, the inspectors presented the results of the radiation safety inspections to Mr. J. Kowalewski, Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. On December 1, 2010, the NRC conducted a teleconference with Mr. R. Murillo, Acting Nuclear Safety Assurance Director, and other members of the licensing staff and asked for additional information related to items of concerns discussed in Section 2RS02. On January 7, 2011, the inspectors informed Mr. J. Pollock, Senior Licensing Specialist, by telephone that the issues of concern would be treated as unresolved items, pending a review of information submitted by the licensee. On January 10, 2011, the inspector presented the results of the Selected Issue Follow-up Inspection of quality assurance and quality control issues to Mr. J. Ridgel, Quality Assurance Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. On January 13, 2010, the resident inspectors presented the inspection results to Mr. J. Kowalewski, Vice President, and other members of the licensee's staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section 2

.3 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as noncited violations.

.1 Title 10 CFR 50.47(b)(4) requires, in part, that a standard classification and emergency action level scheme is in use by the licensee.

Contrary to the above, on October 13, 2010, the licensee identified they had not maintained in effect a standard classification and emergency action level scheme. Specifically, the licensee identified seven examples of failures to implement emergency action level AA1-1 for airborne and liquid effluent releases in 2009 and 2010. In each example, the 200-times discharge permit-specific alarm setpoint value could not be read on the associated process radiation monitor because the value was above the monitor's operating range. This finding was more than minor because it impacted the Emergency Preparedness Cornerstone objective attribute of emergency response organization performance and was evaluated as having very low safety significance (Green) because it was a failure to comply with NRC requirements, was associated with a risk-significant planning standard, and was not a functional failure or degraded function of the planning standard. This condition was documented in Condition Reports CR-WF3-2010-6184 and CR-WF3-2010-6387.

.2 Technical Specification 6.8.1.a requires that the licensee shall establish, implement, and maintain the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Contrary to the above, the licensee identified three occasions where

they failed to comply with this technical specification. First, a work instruction governing maintenance on a safety-related emergency feedwater pump contained incorrect guidance. Second, a step in EN-OP-104 to verify operability of all equipment required for a scheduled mode shift was signed as complete, when all required equipment was not operable. Third, the licensee failed to follow EN-LI-102 and properly recognize the reportability of an event. This failure to comply with the technical specification is violation of NRC requirements. This finding was more than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences and since it is similar to Inspection Manual Chapter 0612, Appendix E, example 2g. The violation was evaluated as having very low safety significance (Green) because it did not represent a loss of a system safety function or the loss of a single train for greater than its allowed outage time. This condition was documented in Condition Report CR-WF3-2010-5923 and CR-WF3-2008-2744.

.3 Procedure,

EN-QV-111, "Training and Certification of Inspection/Verification and Examination Personnel," Section 4.0 [4](i), requires that the Entergy corporate ANSI Level III inspector shall perform periodic (annual) surveillances of quality control inspection activities to ensure that the program is being adequately implemented and maintained. Contrary to the above, no surveillances of quality control inspection activities were performed for any Entergy site during calendar year 2008. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, so it was evaluated using the qualitative criteria listed in Table 4.1. This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist. This issue was entered into the licensee's corrective action program as CR-HQN-2009-00111.

.4 Title 10 CFR 50.120(b)(2), states in part that training programs must be derived from a systems approach to training as defined in 10 CFR 55.4.

Systems approach to training includes five elements, one of which is evaluation of trainee mastery of objectives. Written examinations are utilized by the licensee to evaluate Auxiliary Operator mastery of objectives. Contrary to the above, the licensee failed to properly implement the evaluation element of their systems approach to training. Specifically, an exam compromise was discovered during Cycle 1 of the 2009 Auxiliary Operator Requalification Cycle. This compromise invalidated subsequent auxiliary operator requalification exams during the cycle due to some commonality between weekly exams.

The issue was placed into the licensee's corrective action program as Condition Report CR-WF3-2009-1077. The licensee conducted a root cause evaluation which identified unclear guidance regarding acceptable forms of information sharing during a requalification exam cycle. Upon discovery of the exam compromise, the licensee re-examined all affected Auxiliary Operators with newly written exams to validate the requalification exam cycle. Extent of conditions determined the exam issue was isolated to the non-licensed operator program and was not evident within other accredited programs. The violation was determined to be of very low safety significance because all Auxiliary Operators were administered a valid exam during the requalification cycle and the extent of conditions was limited to the non-licensed operator community.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

J. Kowalewski, Site Vice President

J. Abisamra, Echelon Chief Engineer

C. Arnone, General Manager, Plant Operations

C. Alday, Manager, System Engineering

S. Beagles, Echelon Manager of Fleet Operations

D. Becker, Technical Specialist IV, Programs and Components

C. Becnel, ECP Coordinator

E. Begley, Senior Engineer, Programs and Components

D. Boan, Supervisor, Radiation Protection

E. Brauner, Supervisor, System Engineering

J. Brawly, ALARA Supervisor, Radiation Protection

B. Briner, Technical Specialist IV, Programs and Components

A. Buford, Engineer II, System Engineering

R. Byrd, Echelon Sr. Staff Engineer

K. Christian, Director, Nuclear Safety Assurance

K. Cook, Manager, Operations

J. Dent, Echelon General Manager Plant Operations, Fleet Operations Support

C. England, Manager, Radiation Protection

G. Fey, Manager, Emergency Preparedness

B. Ford Echelon Sr. Manager, Nuclear Safety and Licensing

C. Fugate, Assistant Manager, Operations

J. Hashim, Senior Engineer, Programs and Components

W. Hardin, Licensing, Specialist

E. Harris, Echelon, QA Manager

M. Haydel, Supervisor, Programs and Components

J. Hornsby, Manager, Chemistry

D. Jacobs, Echelon Sr. Vice President of Planning, Development and Oversight

J. Kowalewski, Vice President of Operations

H. Landeche, Jr., Senior Technician, Instruments and Controls

B. Lanka, Manager, Design Engineering

J. Lewis, Senior Project Manager

B. Lindsey, Manager, Maintenance

M. Mason, Senior Licensing Specialist, Licensing

J. McCann, White Plains Vice President of Nuclear Safety, Emergency Preparedness and Licensing

W. McKinney, Manager, Corrective Action and Assessments

P. Morris, Echelon Manager of Administrative Services

R. Murillo, Acting Director, Nuclear Safety Assurance

K. Nichols, Director, Engineering

T. Palmisano, Echelon Vice President of Oversight

R. Perry, Senior Emergency Planner

Attachment

A. Piluti, Manager, Radiation Protection

J. Pollack, Engineer, Licensing

R. Putnam, Acting Director, Engineering

J. Frick, Manager, Plant Security

W. Renz, Director, Emergency Planning, Entergy South

J. Ridgel, Quality Assurance Manager

R. Seemann, Supervisor, Reactor Engineering

J. Solaski, Auditor, Quality Assurance

W. Steelman, Manager, Licensing

T. Tankersly, Echelon Director of Oversight

E. Weinkam, White Plains Sr. Manager of Nuclear Safety and Licensing

J. Williams, Senior Licensing Specialist, Licensing

NRC Personnel

M. Davis, Senior Resident Inspector

D. Overland, Resident Inspector

M. Ashley, Office of Nuclear Reactor Regulation

K. Fuller, Region IV

M. Gray, Region I

J. Geissner, Region III

N. Hilton, Office of Enforcement

D. Holody, Region I

D. Jackson, Region I

W. Jones, Region IV

R. Kellar, Region IV

M. Marsh, Office of General Counsel

M. McLaughlin, Region I

M. Murphy, Office of Nuclear Reactor Regulation

C. Schulten, Office of Nuclear reactor Regulation

D. Thatcher, Office of Nuclear Reactor Regulation

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000382/20100005-01 URI Foreign Material Exclusion Issue associated with the Condensate Storage Pool Gooseneck Vent (1R06)

05000382/20100005-02 URI Removal of Radioactivity from the Steam Generators (2RS02)

05000382/20100005-03 URI Leakage from the Reactor Coolant Pump Seals (2RS02)

Opened and Closed

05000382/20100005-04 NCV Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters (4OA2)

05000382/20100005-05 NCV Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program (4OA2)

Closed

05000382/LER-2009-001-00 LER Mode Change with an Inoperable Emergency Feedwater Pump (4OA3)

05000382/LER-2009-004-00 LER Condition Prohibited by Technical Specification with Log Power Channel Inoperable (4OA3)

Attachment

Section 1R01: Adverse Weather Protection Procedures/Documents

Number Title Revision

OP-901-521 Off Normal Procedure Severe Weather and Flooding 303

G-202 Piping and Valve Stem, Wall & Floor Penetration Details 10 G-580 Nuclear Plant Island Structure Flood Wall Penetrations - Sheets 1 through 3 Plans 2 G-580 Nuclear Plant Island Structure Flood Wall Penetrations - Sheets 4 Plan 4 B-168 Typical

RB-1,

RB-2,

RB-3, &

RB-4 Wall/Floor - Rad Flex Boot Seal - Vendor Drawing 3

Section 1R04: Equipment Alignment Procedures/Documents Number Title Revision

SD-CHW Essential Chilled Water System Description 5

OP-002-004 Chilled Water System 303

OP-002-001 Auxiliary Component Cooling Water 302

SD-CC Component Cooling Water 7

Section 1R05: Fire Protection Procedures/Documents Number Title Revision

UNT-005-013 Fire Protection Program 11

OP-009-004

Fire Protection

307

MM-007-010

Fire Extinguisher Inspection and Replacement

304

FP-001-015

Fire Protection System Impairments

303

FP-001-018 Pre-Fire Strategies 9

OP-903-060 Fire Hose Station Inspection 8 G-1367 Reactor Bldg. & Wing Area Plan El. -35 & El. -4 1 G-1359 Reactor Auxiliary Bldg. Plan El. +21 2

Attachment

Condition Reports

CR-WF3-2010-6889 CR-WF3-2010-6910

Section 1R06: Flood Protection Procedures/Documents Number Title Revision

OP-003-004 Normal Operating Procedure Condensate Makeup 304 G-134 General Arrangement Reactor Auxiliary Bldg. Plan EL +46.00' 34 G-135 General Arrangement Reactor Auxiliary Bldg. Plan EL +21.00' 30 G-136 General Arrangement Reactor Auxiliary Bldg. Plan EL +4.00' 35 G-137 General Arrangement Reactor Auxiliary Bldg. Plan EL -35.00' 27 G-138 General Arrangement Reactor Auxiliary Bldg. Section - Sheet 1 20

MNQ3-5 Flooding Analysis Outside Containment 4

Condition Reports

CR-WF3-2010-7478

CR-WF3-2011-0139 CR-WF3-2011-0140

Section 1R07: Heat Exchangers Procedures/Documents Number Title Revision

EV-DC-316 Heat Exchanger Program 2

PE-001-015 Generic Letter 89-13 Test Basis 5

PE-004-029 SDC Heat Exchanger B Performance Test 0

ER-W3-2001-1125 CCW Monitoring Plan 1

Work Orders

5001002501

Section 1R11: Licensed Operator Requalification Program

Attachment Procedures/Documents Number Title Revision

OP-902-000 Standard Post Trip Actions 10

OP-901-212 Rapid Plant Power Reduction 3

OP-902-007 Steam Generator Tube Rupture Recovery Procedure 12 E-135 Simulator Scenario 8/5/2010

Section 1R12: Maintenance Effectiveness Procedures/Documents Number Title Revision

EN-DC-203 Maintenance Rule Program 1

EN-DC-204 Maintenance Rule Scope and Basis 2

OP-003-014 Control Room Heating and Ventilation (HVC)

MI-005-532 Normal Control room Train A or B Differential Loop check and Calibration HVCIP5068 A or HVCIP5068 B 3 HVCC Maintenance Rule Table for Waterford 3 Steam Electric Station - Scoping Document

1 PM Basis Template EN - I&C - Electronic Circuit Cards 3

Condition Reports

CR-WF3-2009-2538

CR-WF3-2009-6500

CR-WF3-2010-2584

CR-WF3-2010-4742

CR-WF3-2009-3840

CR-WF3-2010-0313

CR-WF3-2010-2736

CR-WF3-2010-4766

CR-WF3-2009-3918

CR-WF3-2010-1512

CR-WF3-2010-3005

CR-WF3-2010-4945

CR-WF3-2009-5568

CR-WF3-2010-2058

CR-WF3-2010-3519

CR-WF3-2010-5964

CR-WF3-2010-5635

Section 1R13: Maintenance Risk Assessment and Emergent Work Controls Procedures/Documents Number Title Revision

EN-WM-101

On-line Work Management Process

6

OI-037-000

Operations' Risk Assessment Guideline

2

EN-DC-151 PSA Maintenance and Updates 1

PRA-W3-01-001 Waterford 3 PRA Summary Report 0

Attachment

PRA-W3-01-001S10 WF3 Equipment Out Of Service (EOOS) Monitor Work Package 3

Condition Reports

CR-WF3-2010-6763 CR-HQN-1010-1315

Section 1R15: Operability Evaluations Condition Reports

CR-WF3-2010-6106

CR-WF3-2010-6363

CR-WF3-2010-6364

CR-WF3-2010-7346

Work Orders

00252983

52037421

00239458

00223866

Procedures/Documents Number Title Revision

EN-OP-104

Operability Determination Process

4

EN-WM-101

On-Line Work Management Process

6

OI-037-000

Operations Risk Management Guideline

300

OP-100-010

Equipment Out of Service

303

W2.502

Configuration Risk Management Program Implementation

0

OP-100-014 Technical Specification and Technical Requirements Compliance 307 W3-DBD-014 Design Basis Documentation DBD No.6 Main Steam 0

OP-006-008 Transformer Operation 301

Section 1R18: Plant Modifications Procedures/Documents Number Title Revision

EN-DC-136 Temporary Modifications 5

EC-9720

SI-401 A/B Timing 0

EC-76609 Refueling Machine Jumper 0

Attachment Work Orders

00150520 00164192

Section 1R19: Post Maintenance Testing Procedures/Documents Number Title Revision

MM-006-021 Charging Pump Maintenance 11

MM-001-069 Shaft Alignment 4

MM-006-120 Valve Packing program 1

OP-903-003 Charging Pump Operability Check 302

UNT-005-007 Plant Lubrication Procedure 303

OP-903-121 Quarterly IST Valve Tests

12

Work Orders

52246139

52247200

00229103

00123916

00233909

52248947

00250779

00253324

00253292

52275908

00258556

00256250 52260501

Section 1R22: Surveillance Testing Procedures/Documents Number Title Revision

OP-903-063 Chilled Water Pump Operability Verification 302

OP-903-050 Component Cooling Water and Auxiliary Component Cooling Water Pump and Valve Operability Test 24

OP-903-121 Safety Systems Quarterly IST Valve Tests

12

Work Orders

52264045

52279344

52260501

Section 2RS01: Radiological Hazard Assessment and Exposure Controls Procedures/Documents Attachment Number Title Revision

EN-RP-100 Radiation Worker Expectations 5

EN-RP-101 Access Control for Radiologically Controlled Areas 5

EN-RP-102 Radiological Control 2

EN-RP-105 Radiological Work Permits 9

EN-RP-108 Radiation Protection Posting 9

EN-RP-121 Radioactive Material Control 6

EN-RP-131 Air Sampling 8

EN-RP-202 Personnel Monitoring 7

HP-002-201 Radiological Survey Techniques and Frequencies 302

Condition Reports

CR-WF3-2009-05878

CR-WF3-2009-06933

CR-WF3-2009-06939

CR-WF3-2009-06944

CR-WF3-2009-07053

CR-WF3-2010-07105

CR-WF3-2009-07106

CR-WF3-2010-04609

CR-WF3-2010-05443

CR-WF3-2010-05458

Radiation Work Permits

20101005 General Inspections and Tours

Audits, Self-Assessments, and Surveillances Number Title Date

QS-2010-W3-04 QA Follow-up Surveillance of Radiation Protection/Radwaste Audit March 31, 2010

Miscellaneous

Document Number Title Date

Refuel Outage 16 Radiation Protection Report

Section 2RS02: Occupational ALARA Planning and Controls Procedures/Documents Number Title Revision Attachment

EN-RP-110 ALARA Program 7

EN-RP-110-01 ALARA Initiative Referrals 0

EN-RP-110-02 Elemental Cobalt Sampling 0

EN-RP-141 Job Coverage 5

N-RP-143 Source Control 7

EN-FAP-RP-001 Corporate ALARA Committee 1

HP-001-114 Control of Temporary Shielding 11

Condition Reports

CR-WF3-2009-05886

CR-WF3-2009-06033

CR-WF3-2009-07262

CR-WF3-2010-01745

Radiation Work Permits

20090508 Inspect/Rework RCP Motors 1B, 2A, 2B to include support work in shrouds

20090513 RCP 1A Motor and Driver Mount Removal and Replacement

20090600 HP Surveys/Roving Job Coverage in the Reactor Containment Building

20090606 Perform Minor Maintenance activities, walkdowns, surveillances, and inspections, NO work on valves, RCP motors/pumps, or the Reactor Head

20090610 Erect/Dismantle Scaffolding in the Reactor Containment Building

20090618 Remove/Replace Insulation in the Reactor Containment Building

20090702 Disassembly of Reactor Head and Associated Work Activities

20090705 Reassembly of Reactor Head and Associated Work Activities including Staging/Destaging of Equipment

20090707 Remove/Replace the ICI guide tubes (Thimbles)

20090708 ICI Removal/Installation to include cut up of ICIs and work on ICI Equipment

Audits, Self-Assessments, and Surveillances Attachment Number Title Date

QA-14/15-2009-W3-1 Quality Assurance Audit Report September 28, 2009

Quality Oversight Observations of Radiation Protection November 30, 2009

LO-WLO-2010-00025-CA-00001 RP Occupational Radiation Safety/ALARA Pre-NRC Inspection February 25, 2010

Miscellaneous

Document Number Title Date

2009-2013 WF3 Five Year ALARA Plan April 27, 2009

Section 4OA1: Performance Indicator Procedures/Documents Number Title Revision

NEI 99-02 Regulatory Assessment Performance Indicator Guideline 6

EN-LI-114 Performance Indicator Process 4

EN-EP-201 Performance Indicators 9, 10

EP-001-001 Recognition and Classification of Emergency Conditions 24, 25

EP-002-010 Notifications and Communications 303, 304

EP-002-052 Protective Action Guidelines 20, 21

ECH-NE-09-00036 Waterford 3 MSPI Basis Document 2

EN-FAP-RP-002 Radiation Protection Performance Indicator Program 0

Section 4OA2: Problem Identification and Resolution Procedures/Documents Number Title Revision

EN-LI-102

Corrective Action Process

16

EN-LI-118 Root Cause Analysis Process 12

Condition Reports

Attachment

CR-WF3-2010-3372

CR-WF3-2010-6563

CR-WF3-2010-6628

CR-WF3-2010-3780

CR-WF3-2010-4025

CR-WF3-2010-6571

CR-WF3-2010-6699

CR-WF3-2010-4117

CR-WF3-2010-4051

CR-WF3-2010-6595

CR-WF3-2010-6718

CR-WF3-2010-4289

CR-WF3-2010-6424

CR-WF3-2010-6620

CR-WF3-2009-00802

CR-WF3-2009-00851

CR-WF3-2010-2278

Work Orders

00212831

00232563

00256250

Documentation for Entergy Wide Inspection Procedures/Documents Number Title Revision/DateEN-LI-121 Entergy Trending Process Rev 8

EN-MA-102 Inspection Program Rev 3 and 4

EN-QV-100 Conduct of Nuclear Oversight Rev 4

EN-QV-109

EN-QV-109-02 Audit Process Audit Process Guidance

Rev 16

Rev 0

EN-QV-111 Training and Certification of Inspection/Verification and Examination Personnel

Rev 8

EN-QV-117 Oversight Training Program Rev 9

EN-QV-119 Corrective Action Requests, Supplier Stop Work Orders, and Recommendations

Rev 6

EN-QV-123 Supplier Audits/Surveys Rev 3

EN-QV-128 Assessments of Nuclear Oversight? Rev 2

EN-QV-129 Vulnerability Review Process Rev 1

Attachment Technical Specifications Section Waterford Unit 3 6.3

Unit Staff Qualifications Arkansas Nuclear One -1 5.3

Unit Staff Qualifications Arkansas Nuclear One -2 6.3

Unit Staff Qualifications Grand Gulf 5.3

Unit Staff Qualifications Indian Point 2

5.3

Unit Staff Qualifications Indian Point 3

5.3

Unit Staff Qualifications River Bend 5.3

Plant Staff Qualifications Vermont Yankee 5.3

Plant Staff Qualifications James A. Fitzpatrick 5.3

Unit Staff Qualifications Palisades Nuclear Plant 5.3

Unit Staff Qualifications Pilgrim Nuclear Power Station 6.2

Unit Staff Qualifications

Condition Reports

CR-ANO-1-2009-02330

CR-ANO-C-2010-01503

CR-ANO-1-2010-00743

CR-ANO-C-2009-01884

CR-ANO-1-2010-01724

CR-ANO-1-2010-01080

CR-ANO-C-2009-02608

CR-ANO-1-2010-01182

CR-ANO-1-2010-00719

CR-ANO-2-2010-00028

CR-JAF-2008-03648

CR-JAF-2009-04592

CR-JAF-2010-03280

CR-HQN-2010-00111

CR-HQN-2009-01188

CR-HQN-2010-00415

CR-HQN-2009-00178

CR-HQN-2009-01197

CR-HQN-2010-00333

CR-HQN-2009-01083

CR-HQN-2010-00013

CR-HQN-2010-00123

CR-HQN-2009-01084

CR-HQN-2010-00386

CR-HQN-2010-00109

CR-HQN-2009-01085

CR-HQN-2010-00571

CR-HQN-2010-00068

CR-HQN-2009-01091

CR-HQN-2010-00593

CR-HQN-2010-00063

CR-HQN-2009-01093

CR-HQN-2010-00515

CR-HQN-2010-00045

CR-HQN-2009-01096

CR-HQN-2010-00550

CR-HQN-2010-00060

CR-HQN-2009-01140

CR-HQN-2010-00511

CR-HQN-2009-01198

CR-HQN-2009-01150

CR-HQN-2010-00510

CR-HQN-2009-01194

CR-HQN-2009-01169

CR-HQN-2010-00475

CR-HQN-2010-00594

CR-HQN-2009-01170

CR-HQN-2010-00499

CR-HQN-2009-01171

CR-HQN-2009-01184

CR-HQN-2010-00338

CR-HQN-2009-01153

CR-IP2-2010-04085

CR-IP3-2009-04917

CR-IP2-2009-05393

CR-IP3-2010-01740

CR-IP3-2009-04920

CR-IP2-2009-05399

CR-IP2-2010-03985

CR-IP3-2009-04897

CR-IP2-2009-05400

CR-IP2-2010-03986

CR-IP2-2009-05404

CR-IP2-2009-05389

CR-IP2-2010-03988

CR-IP2-2009-05409

CR-IP2-2009-05349

CR-IP2-2010-03984

CR-IP3-2009-04868

CR-IP2-2009-05348

CR-IP3-2009-04903

CR-IP3-2009-04883

CR-IP2-2009-05321

CR-IP3-2009-04905

CR-IP3-2009-04884

Attachment

CR-PLP-2009-04108

CR-PLP-2010-02288

CR-PLP-2009-05909

CR-PLP-2009-05613

CR-PLP-2010-02290

CR-PLP-2010-02012

CR-PLP-2009-05918

CR-PLP-2009-05942

CR-PLP-2009-05897

CR-PLP-2009-05908

CR-PNP-2009-01798

CR-PNP-2008-03922

CR-PNP-2009-05303

CR-PNP-2009-02059

CR-PNP-2009-05359

CR-PNP-2009-05297

CR-PNP-2009-02255

CR-PNP-2010-00015

CR-PNP-2010-02124

CR-PNP-2008-00916

CR-RBS-2008-04685

CR-RBS-2010-01472

CR-RBS-2010-00006

CR-RBS-2009-05041

CR-RBS-2010-02033

CR-RBS-2009-06472

CR-RBS-2009-06123

CR-RBS-2010-00200

CR-RBS-2009-06495

CR-RBS-2009-06446

CR-RBS-2010-00221

CR-RBS-2009-06456

CR-RBS-2009-06451

CR-RBS-2010-00278

CR-RBS-2009-06450

CR-RBS-2009-06471

CR-RBS-2010-00088

CR-RBS-2009-06452

CR-RBS-2009-06473

CR-RBS-2010-00011

CR-RBS-2009-06158

CR-RBS-2009-06490

CR-RBS-2009-06520

CR-RBS-2009-06209

CR-RBS-2010-00044

CR-RBS-2009-06539

CR-RBS-2009-06449

CR-WF3-2010-01198

CR-WF3-2010-00284

CR-WF3-2009-07711

CR-WF3-2010-01356

CR-WF3-2009-07713

CR-WF3-2010-02629

CR-WF3-2010-00746

CR-VTY-2009-04496

CR-VTY-2010-04432

CR-VTY-2010-04496

CR-VTY-2010-01479

CR-VTY-2010-04434

CR-VTY-2010-00070

CR-VTY-2010-02759

CR-GGN-2010-04140

CR-GGN-2010-02135

CR-GGS-2009-06921

CR-GGN-2010-02730

CR-GGN-2010-02382

CR-GGS-2009-06922

CR-GGN-2010-04178

CR-GGN-2010-02902

CR-GGS-2009-06923

CR-GGN-2010-04101

CR-GGN-2010-00590

CR-GGS-2009-06927

CR-GGN-2010-04092

CR-GGN-2010-01247

CR-GGS-2009-06806

CR-GGN-2010-03674

CR-GGN-2010-01252

CR-GGN-2010-00164

CR-GGN-2010-03721

CR-GGN-2009-06575

CR-GGN-2009-06904

CR-GGN-2010-03900

CR-GGS-2009-06907

CR-GGN-2009-06910

CR-GGN-2010-03451

CR-GGS-2009-06920

CR-GGN-2009-06505

CR-GGN-2010-03492

Attachment

Miscellaneous Documents

Number Title Revision/DateEOI Letter

ENOC-10-00002 Response to Request for Information, Revision 1 1/8/10 EOI Letter

ENOC-09-00037 Response to Request for Information

11/30/10 QAPM Entergy Quality Assurance Program Manual

0 through 20 Regulatory Guide 1.8 Personnel Selection and Training 1 ANSI/ANS 3.1-1978

American National Standard for Selection and Training of Nuclear Power Plant Personnel 1978 ANSI N18.1-1971

American National Standard for Selection and Training of Nuclear Power Plant Personnel 1971 NRC SER NRC Safety Evaluation Report, "Entergy Operations, Inc. Quality Assurance Program Consolidation"

11/6/98 Technical Specification

Unit Staff Qualifications various 5.3.1

CEO2009-00195

EOI Letter

BVY 03-12

CIN-2003/00059

EOI Letter No.

CNRO-2003-013

EOI Letter No.

CEXO-2003/164

EOI Letter NO.

CNRO-2002/027

10

CFR 50.59 Review Form Personnel Change Planning Checklist/Forms for QA Manager Candidates

Corporate ANSI Level III Surveillance of VY Maintenance Inspection Program (VTY)

Vermont Yankee Nuclear Power Station, Docket No. 50-271 Annual Submittal of QAP Changes (VTY)

Vermont Yankee, 10 CFR Part 50.54(a)(3) Change Review Forms for QAPM

Entergy Quality Assurance Program Manual, Rev. 8 (VTY)

Issuance of Entergy Quality Assurance Program Manual

(QAPM) Revision 8 (VTY)

Entergy Quality Assurance Program Manual, Revision 7 (PNPS)

July 2007

12/15/2009

02/05/2003

04/24/2002

Rev 8 (VTY)

04/24/2003

04/24/2003

04/25/2002

Attachment

ENO Letter No. 1.2.02-067

EN-QV-104

Attachment 9.1

ENOC Letter NO. 07-0020

AP-20.06, Attachment 1

MCM-4.1 Attachment 4.1

AP-20.09 Attachment 1

Entergy Letter

JLIC-02-017

ENO Letter 1.2.02-060

Entergy Letter

CNRO-2002-027

10

CFR 50.54(a)

Evaluation

ENO Letter 1.2.02-060

ENO Meeting Summary

Entergy QA Program Manual, Revision 7 (PNPS)

Entergy QA Program Manual, Revision 7 (PNPS)

Independent Spent Fuel Storage Installation Entergy QA Program Manual Change Review Form 50.54(a) Parts 1,2 and 3 (PLP)

Entergy QA Program Manual, Revision 16, Annual Report 10

CFR 50.54(a)(3) and10

CFR 72.140(d) (PLP)

FSAR Change Request Form, Relocate QA Program from Chapter 17 to Entergy QAPM (JAF)

Nuclear Engineering 10

CFR 50.59 Screening Form (JAF)

Process Applicability Screening - Relocate QA Program From FSAR Ch. 17 to Entergy QAPM (JAF)

Cross Reference of QAPM commitments to Implementing procedures at JAF

Adaptation of Entergy Common QAPM, Revision 7 (JAF)

Entergy QA Program Manual, Revision 7 (JAF)

QA Program Change/Prior Approval Determination - Part A (IP3)

Adaptation of Entergy Common QAPM, Revision 7, (IP2 and IP3)

Development of Common QA Manual for northern Entergy Sites and Entergy Nuclear Generating Company Plants

05/02/2002

07/30/2002

04.05/2007

04/15/2007

05/06/2002

04/03/2002

04/01/2002

04/02/2002

06/21/2002

04/25/2002

05/06/2002

06/21/2002

11/30/2001

Engineering Changes/Maintenance Work Orders

ANO-EC-07032

RBS-EC-00893

RBS-EC-70734

GGN-EC-01450

PLP-EC-05885

ANO-EC-02886

RBS-EC-02692

GGN-EC-00085

GGN-EC-01452

PLP-EC-09121

ANO-EC-03069

RBS-EC-03275

GGN-EC-00224

GGN-EC-02048

PLP-EC-12392

ANO-EC-04461

RBS-EC-03643

GGN-EC-02048

GGN-EC-02065

PLP-EC-14181

ANO-EC-08043

RBS-EC-03850

GGN-EC-02058

GGN-EC-13326

PLP-EC-18042

ANO-EC-00608

RBS-EC-03275

GGN-EC-02065

GGN-EC-13354

PLP-EC-06553

Attachment

WF3-EC-15451

RBS-EC-05932

GGN-EC-02107

GGN-EC-13355

PLP-EC-12731

WF3-EC-10706

RBS-EC-06947

GGN-EC-02110 ANO U-1

EC 01039

WF3-EC-01830

RBS-EC-07239

GGN-EC-02201 ANO U-1

EC 05808

WF3-EC-07960

RBS-EC-08504

GGN-EC-02784 ANO U-1

EC 13153

WF3-EC-01166

RBS-EC-12204

GGN-EC-04538 ANO U-1

EC 00380

WF3-EC-09046

RBS-EC-13128

GGN-EC-06299 ANO U-1

EC 05054

WF3-EC-00935

RBS-EC-16451

GGN-EC-06301 ANO U-1

EC 05388

WF3-EC-01166

RBS-EC-70752

GGN-EC-07471 ANO U-1

EC 06241

WF3-EC-01396

RBS-EC-07368

GGN-EC-07716 ANO U-1

EC 07032

WF3-EC-01782

RBS-EC-03852

GGN-EC-06875 ANO U-1

EC 13224

WF3-EC-03013

RBS-EC-03853

GGN-EC-06039

WF3-EC-844881

WF3-EC-11284

RBS-EC-03975

GGN-EC-06086

WF3-EC-05854

WF3-EC-13981

RBS-EC-70733

GGN-EC-00494

VYT-EC-03138

Audit Reports/Surveillances

Corporate ANSI Level III Surveillance of VY Inspection Program PNP Pre-NIEP 2009 Report

PNP Pre-NIEP 2010 VY Pre-NIEP 2007

LO-VTYLO-2007-00029 Palisades Pre-NIEP 2009 Palisades 2008 Pre-NIEP Report JAF Pre-NIEP August 2007

IPEC Pre-NIEP 2009

PEC 2008 Pre- NIEP Assessment GGNS Pre-NIEP Report final May 2008 GGNS Pre-NIEP 2009 ANO Pre-NIEP 2010

WF3 Pre-NIEP 2007 W3 CEO2008-00026

QA-13-2009-PLP-01 PLP NIEP 2009

QA-13-2009-GGNS-1 GGNS NIEP 2009

QA-13-2007-VY-1 NIEP AUDIT REPORT NIEP - River Bend - 2007

JAF

QA 2008 NIEP Report IPEC 2009 NIEP Report WF3 NIEP 2008

QA-10-2006-VY-1 Maintenance

QA-10-2006-RBS-1 Maintenance

QA-10-2006-JAF-1 Maintenance

QA-10-2006-PNP-1Maintenance

QA-10-2006-IP-1 Maintenance

QA-10-2006-GGNS-1 Maintenance

QA-10-2006-ANO-1 Maintenance

QA-10-2006-WF3-1 Maintenance

QS-2010-PLP-017 PLP QC Inspection Program

QS-2010-GGNS-011 GGNS QC Inspection Program

QS-2010-ECH-008 ANSI Level III of IPEC

Attachment

QS-2010-ECH-007 Review of EOC for QC Inspection Point Selection

QS-2010-ECH-006 Review of Fleet Interim Actions

QS-2010-ECH-002 ANSI Level III of PNP

QS-2010-ECH-001 ANSI Level III of GGNS

QS-2009-VY-004 VY Inspection Program

QS-2009-VY-020 VY Maintenance Inspection Program

QS-2009-ANO-006 Corporate ANSI Level III of ANO

QS-2008-VY-004 Peer Inspector Qualification Documentation

QS-2010-PNPS-019 PNP Inspection Program

QA-10-2008-VY-1 Maintenance

QA-10-2008-RBS-1 Maintenance

QA-10-2008-PNP-1 Maintenance

QA-10-2008-PLP-1 Maintenance

QA-10-2008-JAF-1 Maintenance

QA-10-2008-IP-1 Maintenance

QA-10-2008-GGNS-1 Maintenance

QA-10-2008-ANO-1 Maintenance

QA-10-2008-WF3-1 Maintenance Corporate ANSI Level III Surveillance of VY Inspection Program PNP Pre-NIEP 2009 Report PNP Pre-NIEP 2010 VY Pre-NIEP 2007

LO-VTYLO-2007-00029 Palisades Pre-NIEP 2009

Palisades 2008 Pre-NIEP Report JAF Pre-NIEP August 2007 IPEC Pre-NIEP 2009 IPEC 2008 Pre- NIEP Assessment GGNS Pre-NIEP Report final May 2008

GGNS Pre-NIEP 2009 ANO Pre-NIEP 2010 WF3 Pre-NIEP 2007 W3 CEO2008-00026

QA-13-2009-PLP-01 PLP NIEP 2009

QA-13-2009-GGNS-1 GGNS NIEP 2009

QA-13-2007-VY-1 NIEP AUDIT REPORT NIEP - River Bend - 2007 JAF

QA 2008 NIEP Report IPEC 2009 NIEP Report WF3 NIEP 2008

QA-10-2006-VY-1 Maintenance

QA-10-2006-RBS-1 Maintenance

QA-10-2006-JAF-1 Maintenance

QA-10-2006-PNP-1Maintenance

QA-10-2006-IP-1 Maintenance

QA-10-2006-GGNS-1 Maintenance

QA-10-2006-ANO-1 Maintenance

QA-10-2006-WF3-1 Maintenance

QS-2010-PLP-017 PLP QC Inspection Program Attachment

QS-2010-GGNS-011 GGNS QC Inspection Program

QS-2010-ECH-008 ANSI Level III of IPEC

QS-2010-ECH-007 Review of EOC for QC Inspection Point Selection

QS-2010-ECH-006 Review of Fleet Interim Actions

QS-2010-ECH-002 ANSI Level III of PNP

QS-2010-ECH-001 ANSI Level III of GGNS

QS-2009-VY-004 VY Inspection Program

QS-2009-VY-020 VY Maintenance Inspection Program

QS-2009-ANO-006 Corporate ANSI Level III of ANO

QS-2008-VY-004 Peer Inspector Qualification Documentation

QS-2010-PNPS-019 PNP Inspection Program

QA-10-2008-VY-1 Maintenance

QA-10-2008-RBS-1 Maintenance

QA-10-2008-PNP-1 Maintenance

QA-10-2008-PLP-1 Maintenance

QA-10-2008-JAF-1 Maintenance

QA-10-2008-IP-1 Maintenance

QA-10-2008-GGNS-1 Maintenance

QA-10-2008-ANO-1 Maintenance

QA-10-2008-WF3-1 Maintenance

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion

Condition Reports

CR-WF3-2010-5923

CR-WF3-2009-0975

CR-WF3-2008-2744

CR-WF3-2010-6044

Work Orders

00118868

00180656

Procedures/Documents Number Title Revision

OP-010-003 Plant Startup 306

MI-005-425 Emergency Feedwater Pump Turbine Governing Control System Calibration 5

Section 4OA7: Licensee-Identified Violations

Miscellaneous

Number Title Revision/Date Standing Instruction 10-14-2010 October 14, 2010

Liquid Waster Batch Release Permit Attachment LB2009-0100 Procedure

EP-001-001 Recognition and Classification of Emergency Conditions 24

Condition Reports

CR-WF3-2010-6184 CR-WF3-2010-6387