IR 05000382/2005005

Nrc Integrated Inspection Report 05000382-05-005

ML060320708
Person / Time
Site:	Waterford
Issue date:	01/31/2006
From:	Graves D NRC/RGN-IV/DRP/RPB-E
To:	Venable J Entergy Operations
References
IR-05-005
Download: ML060320708 (33)

Text

ary 31, 2006

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2005005

Dear Mr. Venable:

On December 31, 2005, the NRC completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed report documents the inspection findings which were discussed on January 10, 2006, with Mr. Kevin Walsh, General Manager, Plant Operations, and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.

Based on the results of this inspection, the NRC identified one issue that was evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that a violation was associated with this issue. The violation is being treated as a noncited violation, consistent with Section VI.A of the Enforcement Policy. The finding is described in the subject inspection report. If you contest the subject or severity of a noncited violation, 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, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector 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, its enclosures, and your response will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Entergy Operations, Inc. -2-Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

/RA/

David N. Graves, Chief Project Branch E Division of Reactor Projects Docket: 50-382 License: NPF-38

Enclosure:

NRC Inspection Report 050000382/2005005 w/Attachment: Supplemental Information

REGION IV==

Docket No.: 50-382 License No.: NPF-38 Report No.: 05000382/2005005 Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy. 18 Killona, Louisiana Dates: September 27 through December 31, 2005 Inspectors: M. C. Hay, Senior Resident Inspector G. F. Larkin, Resident Inspector T. F. Stetka, Senior Operations Engineer G. W. Johnston, Senior Operations Engineer P. J. Elkmann, Emergency Preparedness Inspector D. H. Overland, Reactor Inspector B. X. Tindell, Reactor Inspector B. K. Tharakan, Health Physicist Approved By: David N. Graves, Chief, Project Branch E ATTACHMENTS: Supplemental Information Enclosure

SUMMARY OF FINDINGS

IR05000382/2005-005; 09/27/2005-12/31/2005; Waterford Steam Electric Station, Unit 3;

Surveillance Testing The report covered a 3-month period of inspection by resident inspectors along with announced inspections by a regional senior emergency preparedness inspector, two reactor inspectors, a health physicist, and two senior operations engineers. The inspectors identified one Green finding. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 3, dated July 2000.

Cornerstone: Initiating Events

Green.

The inspectors identified a noncited violation for the failure to comply with Technical Specification 3.4.5, Leakage Detection Systems based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection System.

Specifically, the containment fan cooler condensate flow switches were identified to not meet the design requirements for detecting a one gallon per minute reactor coolant system leak. The licensee took prompt corrective actions and entered this issue into the corrective action program.

The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, Significance Determination Process, were used to conclude that the finding was of very low safety significance (Green) because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that the licensee had missed several opportunities to identify this non-conforming condition (Section R22).

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status: The plant began the period on September 27, 2005, at 100 percent power. On November 11, 2005, plant operators manually tripped the reactor due to decreasing main condenser vacuum following a relay failure in the circulating water logic circuitry that resulted in a loss of all running circulating water pumps. On November 14, 2005, operators commenced a reactor startup, synchronized the main turbine generator to the offsite electrical grid, and reached approximately 100 percent reactor power on November 15, 2005. Power remained at that level for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors performed a partial walkdown of the following two systems to verify the operability of redundant or diverse trains and components when safety equipment was inoperable. The inspectors attempted to identify any discrepancies that could impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, walked down control systems components, and verified that selected breakers, valves, and support equipment were in the correct position to support system operation. 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 process.

• On September 26, 2005, the inspectors performed a partial equipment alignment inspection of a critical portion of reactor auxiliary building controlled area ventilation system Train B. The walkdown was conducted using Operating Procedure OP-002-010, "Reactor Auxiliary Building HVAC and Containment Purge," Revision 14.

• On November 8, 2005, the inspectors performed a partial equipment alignment inspection of low-pressure safety injection system Train A. System configuration was assessed using Operating Procedure OP-009-008, "Safety Injection System," Revision 18.

b. Findings

No Findings of significance were identified.

1R05 Fire Protection

.1 Fire Protection - Tours

a. Inspection Scope

The inspectors performed inspections of the six areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that combustibles and ignition sources, were controlled in accordance with the licensees administrative procedures; fire detection and suppression equipment was available for use; that passive fire barriers were maintained in good material condition; and that compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with the licensees fire protection program.

• Fire Zone RAB 2, 18, Roof East and Roof West on October 18, 2005

• Fire Zone RAB 1A, 18, 20, and 21 on October 18, 2005

• Fire Zone RAB 2, 8A, 19, 20, and 32 on October 26, 2005

• Fire Zone Fuel Handling Building +46", +21", +1", and -35" levels on November 28, 2005

• Fire Zone RAB 1A, 1E, 8B, and TGB +15 level on December 12, 2005

• Fire Zone RAB 16, 19, 20, 36, 37, and 39 on December 20, 2005

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

.1 Internal Flooding

a. Inspection Scope

The inspectors performed a semiannual inspection of internal flood protection features in the reactor auxiliary building wing area. The inspection reviewed the design adequacy and use of various safety-related instruments that provide control room indication for area flooding from inadvertent fire protection system sprinkler actuation or high and medium energy line breaks. The inspection included a review of the Updated Final Safety Analysis Report, selected design calculations, and a walkdown of flood protection features in the reactor auxiliary building.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

.1 Licensed Operator Simulator Training

a. Inspection Scope

On October 20, 2005, the inspectors observed one licensed operator simulator training scenario. During the scenario, operators responded to problems associated with an excore nuclear instrument, a component cooling water pump failure, a steam generator tube rupture, along with several additional abnormal equipment conditions occurring.

The simulator training evaluated the operators ability to recognize, diagnose, and respond to abnormal and emergency reactor plant conditions. The inspectors observed and evaluated the following areas:

• Understanding and interpreting annuniciator and alarm signals

• Verifying automatic actions and analyzing plant parameters in abnormal and emergency conditions

• Use and adherence of Technical Specifications

• Communicating as a team and prioritizing actions with attention to detail

• The crew's and evaluator's critiques

• Classifying emergencies and making notifications

b. Findings

No findings of significance were identified.

.2 Biennial Inspection

a. Inspection Scope

To assess the performance effectiveness of the licensed operator requalification program, the inspectors conducted personnel interviews, reviewed both the operating and written examinations, and observed ongoing operating examination activities.

The inspectors interviewed eight personnel, five instructors, two operators, and one training supervisor, to determine their understanding of the policies and practices for administering requalification examinations. The inspectors also reviewed operator performance on the written and operating examinations. These reviews included observations of portions of the operating examination by the inspectors. The operating examinations observed included five job performance measures and five scenarios that

were used in the current biennial requalification cycle. These observations allowed the inspectors to assess the licensee's effectiveness in conducting the operating test to ensure operator mastery of the training program content.

The results of these examinations were reviewed to determine the effectiveness of the licensees appraisal of operator performance and to determine if feedback of performance analyses into the requalification training program was being accomplished.

The inspectors interviewed members of the training department and reviewed minutes of training review group meetings to assess the responsiveness of the licensed operator requalification program in incorporating the lessons learned from both plant and industry events. Examination results were also assessed to determine if they were consistent with the guidance contained in NUREG 1021 and NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process."

Additionally, the inspectors assessed the Waterford plant-referenced simulator for compliance with 10 CFR 55.46, Simulator Facilities. This assessment included the adequacy of the licensees simulation facility for use in operator licensing examinations and for satisfying experience requirements as prescribed by 10 CFR 55.46. The inspectors reviewed a sample of simulator performance test records (transient tests, surveillance tests, and malfunction tests), simulator deficiency report records, and processes for ensuring simulator fidelity commensurate with 10 CFR 55.46. The inspectors reviewed selected simulator deficiency reports generated by the licensee that did not result in changes to the configuration of the simulator to assess the responsiveness of the licensee's simulator configuration management program. The inspectors also interviewed members of the licensees simulator configuration control group as part of this review.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Routine Reviews

a. Inspection Scope

The inspectors reviewed the two samples listed below for items such as:

(1) appropriate work practices;

(2) identifying and addressing common cause failures;

(3) scoping in accordance with 10 CFR 50.65(b) of the maintenance rule;

(4) characterizing reliability issues for performance;

(5) trending key parameters for condition monitoring;

(6) charging unavailability for performance;

(7) classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); and

(8) appropriateness of performance criteria for structures, systems, and components functions classified as (a)(2) and/or appropriateness of goals and corrective actions for structures, systems, and components functions classified as (a)(1).

• Anticipated transient without scram system

• Sump pump system

b. Findings

No findings of significance were identified.

.2 Biennial Maintenance Rule Implementation

a. Inspection Scope

Periodic Evaluation Reviews The inspectors reviewed the licensees overall implementation of the Maintenance Rule, 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. The inspectors reviewed the licensee's Maintenance Rule periodic assessments for the periods of April 2002 to November 2003 and December 2003 to April 2005. The resulting adjustments to the balance of equipment reliability and availability were also evaluated.

The inspectors reviewed systems and functions that had suffered some degraded performance or condition to assess the licensees periodic evaluation activities. The inspectors selected the following six systems for a detailed review:

• Core Protection Calculators

• Instrument Air

• Emergency Diesel Generators

• Reactor Coolant System

• Broad Range Gas Monitor

• Steam Bypass Control For these systems, the inspectors reviewed the use of performance history and operating experience in adjusting preventive maintenance, (a)(1) goals, and (a)(2) performance criteria. The inspectors also reviewed adjustments to the scope of the Maintenance Rule Program and the definitions of availability hours and required available hours.

b. Findings

No findings of significance were identified.

.3 Identification and Resolution of Problems

a. Inspection Scope

The inspectors evaluated the use of the Corrective Action Program within the Maintenance Rule Program. The review was accomplished by the examination of a sample of corrective action documents and work orders. The purpose of the review was to determine that the identification of problems and implementation of corrective actions were acceptable.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the following three activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed.

• On October 6, 2005, during planned surveillance testing on emergency diesel Generator B lockout relays

• On November 1, 2005, during emergent repairs on charging Pump A

• On November 29, 2005, during planned maintenance on start up transformer Train B

b. Findings

No findings of significance were identified.

1R14 Operator Performance During Non-Routine Evolutions and Events

a. Inspection Scope

For the non-routine events described below, the inspectors reviewed operator logs, plant computer data, and strip charts to determine what occurred and how the operators responded, and to determine if the response was in accordance with plant procedures:

• On September 27, 2005, the inspectors observed the site response to a complete loss of all operating circulating water pumps. Operators initiated a manual scram following lowering main condenser vacuum conditions. Lowering main condenser vacuum resulted in loss of main feedwater causing steam generator levels to lower followed by an automatic actuation of the emergency feedwater system.

• On December 8, 2005, the inspectors observed the site response to a main condenser tube rupture resulting in elevated sodium and cation concentrations affecting both steam generators. Operators responded by isolating the affected main condenser waterbox and taking actions to promptly restore chemistry parameters in both steam generators to acceptable concentrations for plant operations.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the technical adequacy of the four operability evaluations listed below to verify that they were sufficient to justify continued operation of a system or component. The inspectors considered that, although equipment was potentially degraded, the operability evaluation provided adequate justification that the equipment could still meet its Technical Specification, Updated Final Safety Analysis Report, and design-bases requirements and that the potential risk increase contributed by the degraded equipment was thoroughly evaluated.

• Operability evaluation addressing the chemical and volume control system following the failure of CVC-218A, charging line to reactor coolant system Loop 1A isolation (Condition Report CR-WF3-2005-4293)

• Operability evaluation addressing emergency diesel generator Train A, unexpected load instability during a 24-hour operability run (Condition Report CR-WF3-2005-4943)

• Operability evaluation addressing divergent pressurizer level control channel indications (Condition Report CR-WF3-2005-4044)

• Operability evaluation addressing intermittent emergency diesel generator fuel injector leakage (Condition Report CR-WF3-2005-3830)

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors reviewed the five postmaintenance tests listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the licensees test procedure to verify that the procedure adequately tested the safety function(s) that may have been affected by the maintenance activity, that the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or design basis documents, and that the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed the test data, to verify that test results adequately demonstrated restoration of the affected safety function(s).

• Control room emergency ventilation Valve HVC-213A on November 26, 2005

• Chemical and volume control system Valve CVC 218A, following emergent work on November 2, 2005

• Chemical and volume control system Pump A, following emergent work on November 2, 2005

• Component cooling water pump Train A, following maintenance on December 9, 2005

• Main feedwater isolation valve number Train A, following maintenance on December 12, 2005

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed or reviewed the following five surveillance tests to ensure the systems were capable of performing their safety function and to assess their operational readiness. Specifically, the inspectors considered whether the following surveillance tests met Technical Specifications, the Updated Final Safety Analysis Report, and licensees procedural requirements:

• Surveillance Procedure OP-903-116, Train B Integrated Emergency Diesel Generator/Engineering Safety Features Test, Revision 9, performed on November 7, 2005. This surveillance verified that a lockout relay will provide diesel generator protection by preventing the diesel from starting concurrent with loss of DC control power, turning gear engaged, emergency stop depressed, or the manual fuel shutdown lever in manual off.

• Surveillance Procedure MI-003-409, Containment Air Cooler Condensate Flow Switches Channel Functional Test, Revision 6, performed on May 24, 2005.

This surveillance verified operability of the containment air cooler condensate flow switch to detect 1 gpm leakage from the reactor coolant system within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• Surveillance Procedure OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification, Revision 14, performed on December 19, 2005. This surveillance verifies the functional capability of the emergency diesel generator Train A to start, load, and run.

• Surveillance Procedure ME-003-302, Emergency Diesel Generator Undervoltage Override and Sequencer Lockout Logic Circuit Testing, Revision 1, performed on October 25, 2005.

• Surveillance Procedure OP-903-120, Containment and Miscellaneous System Quarterly IST Valve Tests, Revision 6, performed on November 28, 2005. This surveillance verified operability of the gaseous waste management Valves GWM-104.

b. Findings

Introduction.

The inspectors identified a Green noncited violation (NCV) for the failure to comply with Technical Specification 3.4.5, Leakage Detection Systems based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection System.

Description.

Waterford 3 Final Safety Analysis Report, Section 5.2.5, Detection of Leakage Through Reactor Coolant Pressure Boundary, states, in part, that the leakage detection systems are consistent with the recommendations of NRC Regulatory Guide 1.45 and these systems are capable of reliably detecting unidentified sources of abnormal leakage as low as one gallon per minute (gpm). Regulatory Guide 1.45, also states that the sensitivity and response time of each leakage detection system employed for unidentified leakage should be adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour.

The inspectors noted that one method utilized to detect unidentified leakage at the facility utilizes containment fan cooler condensate flow switches that monitor the four containment fan cooler pan drains that are each calibrated to alarm at a one gpm flow rate. Based on the licensee typically running multiple fan coolers during plant operation the inspectors noted that a one gpm reactor coolant steam leak could be distributed to all running containment fan coolers resulting in each condensate flow switch seeing less than the one gpm total. During discussions with the licensee the inspectors were informed that no analysis had been performed to assess the adequacy of this design to meet the intent of detecting a one gpm reactor coolant system leak and ultimately the licensee concluded the system was not adequate to meet the intent of the design requirements. This condition had existed since initial plant operations.

The inspectors determined that the licensee had missed several opportunities to identify this non-conforming condition. In 2004, NRC inspectors identified that the containment gaseous radioactivity monitoring system failed to meet the design requirements of Regulatory Guide 1.45 resulting in the licensee declaring this method of detection inoperable. In 2004, the licensee requested and received License Amendment No. 197 that, in part, specifically addressed changes associated with utilizing the containment fan cooler condensate flow switches as a method of leakage detection to satisfy TS 3.4.5. In 2005, the licensee submitted licensee event report (LER) 05000382/2005-002 after identifying that the containment fan cooler condensate flow switches were not being properly calibrated to detect a one gpm flow rate.

Analysis.

The performance deficiency associated with this finding involved the failure to comply with Technical Specification 3.4.5, Leakage Detection Systems based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection

System. The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, Significance Determination Process, were used to conclude that the finding was of very low safety significance (Green). The finding was determined to be of very low safety significance because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that, as previously discussed, the licensee had missed several opportunities to identify this non-conforming condition.

Enforcement.

Technical Specification 3.4.5.1 describes three methods of detecting reactor coolant system leakage and states, in part, that with only two of the required leakage detection systems operable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactivity monitoring system is inoperable; otherwise be in at least Hot Standby within the next six hours and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. As previously stated, the containment air cooler condensate flow switches were inoperable since initial plant startup do to not meeting the design criteria of Regulatory Guide 1.45. Additionally, as previously stated, the inspectors identified that the containment gaseous radioactivity monitoring system failed to meet Regulatory Guide 1.45 design criteria since initial plant startup. These two conditions not being identified by the licensee resulted in the failure to place the plant in a cold shutdown condition as required by Technical Specification 3.4.5.1. The corrective actions to restore compliance included utilizing the containment sump flow monitoring system in addition to the containment fan cooler flow switches to meet the design requirements of Regulatory Guide 1.45. Additionally, the licensee plans to submit changes to Technical Specification 3.4.5.1 in order to clearly define the operability requirements of the leakage detection systems. Because this finding is of very low safety significance and has been entered into the licensees corrective action process as CR-WF3-2005-04267, this violation is being treated as an NCV consistent with Section VI.A of the Enforcement Policy. This violation is identified as NCV 05000382/2005005-01, Inadequate Reactor Coolant System Leakage Detection System.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed Temporary Alteration TA-05-005, Reactor Vessel Head Seal Leakage. The inspectors reviewed the safety screening, design documents, UFSAR, and applicable Technical Specifications to determine that the temporary modification was consistent with the modification documents, drawings, and procedures. The inspectors walked down accessible portions of the affected equipment. The inspectors reviewed the adequacy of post installation tests and test results to confirm that the actual impact of the temporary modification on the permanent system and interfacing systems was adequately verified.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

On October 24, 2005, the inspector observed an operator requalification examination for shift Crew E in the control room simulator. The inspectors observed the training evolution to identify any weaknesses and deficiencies in classification, notification, and Protective Action Requirements development activities. The inspectors evaluated the drill conduct and the adequacy of the licensees critique of performance, with respect to the self-identification of weaknesses and deficiencies. The inspectors reviewed the postdrill critique to ensure that the items independently identified by the inspectors were also identified through the licensees critique and properly entered into the corrective action program.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Occupational Radiation Safety [OS] 2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspector used the requirements in 10 CFR Part 20 and the licensees procedures required by Technical Specifications as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:

• Current 3-year rolling average collective exposure

• Site specific ALARA procedures

• Three work activities of highest exposure significance completed during the last outage

• Intended versus actual work activity doses and the reasons for any inconsistencies

• Integration of ALARA requirements into work procedure and radiation work permit (or radiation exposure permit) documents

• Dose rate reduction activities in work planning

• Source-term control strategy or justifications for not pursuing such exposure reduction initiatives

• Radiation worker and radiation protection technician performance during work activities in radiation areas and high radiation areas

• Self-assessments, audits, and special reports related to the ALARA program since the last inspection

• Corrective action documents related to the ALARA program and follow-up activities such as initial problem identification, characterization, and tracking During this inspection, the inspector completed 6 of the required 15 samples and 4 of the optional samples. For this biennial inspection period, all 15 required samples and 7 optional samples were completed.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)

a. Inspection Scope

This area was inspected to determine the accuracy and operability of radiation monitoring instruments that are used for the protection of occupational workers. The inspector used the requirements in 10 CFR Part 20 and the licensees procedures required by technical specifications as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:

• Calibration of whole body counting equipment and radiation detection instruments utilized for personnel and material release from the radiologically controlled area

b. Findings

No findings of significance were identified.

2PS2 Radioactive Material Processing and Transportation (71122.02)

a. Inspection Scope

This area was inspected to verify that the licensees radioactive material processing and transportation program complies with the requirements of 10 CFR Parts 20, 61, and 71 and Department of Transportation regulations contained in 49 CFR Parts 171-180. The inspector interviewed licensee personnel and reviewed:

• Shipment packages, surveys, labels, markings, placards, vehicle checks, driver instructions, and disposal manifests

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

a. Inspection Scope

The inspector sampled licensee submittals for the performance indicators listed below for the period October 2004 through September 2005. The definitions and guidance of NEI 99-02, Regulatory Assessment Indicator Guideline, Revisions 2 and 3, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of performance indicator data reported during the assessment period.

Licensee performance indicator data were also reviewed against the requirements of Procedures EN-LI-114, Performance Indicator Process, Revision 0, and EN-EP-201, Emergency Planning Performance Indicators, Revision 2.

Emergency Preparedness Cornerstone:

• Drill and Exercise Performance

• Emergency Response Organization Participation

• Alert and Notification System Reliability The inspector reviewed a 100 percent sample of drill and exercise scenarios and licensed operator simulator training sessions, notification forms, and attendance and critique records associated with training sessions, drills, and exercises conducted during the verification period. The inspector reviewed 35 selected emergency responder qualification, training, and drill participation records. The inspector reviewed alert and notification system testing procedures, maintenance records, and a 100 percent sample of siren test records. The inspector also interviewed licensee personnel responsible for collecting and evaluating performance indicator data. The inspector completed 3 samples during this inspection.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Semiannual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, the inspectors performed a review of the licensees corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review focused on repetitive equipment issues, but also considered the results of screening the corrective action program, self-assessment reports, control room logs, quality assurance audits, maintenance rule assessments, and department self-assessments to determine if additional adverse trends existed. The inspectors compared and contrasted their results with the results contained in Entergys latest quarterly trend reports. For those areas where trends were documented in the corrective action program, the inspectors verified that Entergy had corrective actions planned or in place to address the trend. The inspectors also evaluated the corrective actions against Entergys procedural requirements of Procedure LI-102, Corrective Action Program. The inspectors review nominally considered the 6-month period of July through December 2005.

b. Findings and Observations

No findings of significance were identified. The inspectors concluded that, in general, Entergy had adequately identified trends in areas within the scope of this inspection.

.2 Section 2OS2 evaluated the effectiveness of the licensee's problem identification and

resolution processes regarding exposure tracking, higher than planned exposure levels, and radiation worker practices. The inspector reviewed the corrective action documents listed in the attachment against the licensees problem identification and resolution program requirements.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

(Closed) LER 05000382/2005-004-00. Loss of Offsite Power (LOOP) During Hurricane Katrina On August 29, 2005, a LOOP occurred with the plant in Mode 4. At the time of the LOOP, the plant was experiencing tropical storm winds of approximately 48 miles per hour. All loads supplied by the non-safety 6.9 kV busses were lost including the reactor coolant pumps. The two emergency diesel generators started in response to the LOOP conditions and safe shutdown loads were sequenced onto the two safety busses.

Offsite power was available on August 31, 2005. After evaluating grid stability offsite power for Train A was declared operable on September 1, 2005, and offsite power for Train B was declared operable on September 2, 2005. The loss of offsite power

occurred due to transmission system damage caused by Hurricane Katrina. The LER was reviewed by the inspectors and no findings of significance were identified and no violation of NRC requirements occurred. The licensee documented this event in CR-WF3-2005-3872. This LER is closed.

4OA5 Other Activities

.1 Power Uprate (Inspection Procedure 71004)

a. Inspection Scope

The inspectors reviewed a sample of extended power uprate modifications to verify that the modifications were prepared in accordance with the licensing basis and the UFSAR.

Some of the activities that the inspectors reviewed were documented in previous inspection reports including postmaintenance testing of the low steam generator pressure trip bistables and annunciator setpoint modification, a permanent plant modification to revise atmospheric dump valve control loop, a review of emergency diesel Generator A and B jacket water and lube oil heat exchangers performance testing. Additional work not previously documented include a review of power ascension testing to ensure that plant response was as expected, closure tests of the main feedwater isolation valves, a review to verify that the licensee has established a program to adequately monitor and assess flow accelerated corrosion, and a review of the emergency feedwater actuation system control and protective circuitry.

b. Findings

No findings of significance were identified.

.2 (Closed) Unresolved Item (URI) 050000382/2004004-05 Review Revised Procedure and

Re-perform Control Room Evacuation Walkthrough During the August 2003 triennial fire protection inspection, the team opened URI 050000382/2003011-02 concerning the adequacy of control room evacuation Procedure OP-901-502. The inspectors were concerned that the licensee could not meet the critical times in the procedure because it required the operators to diagnose spurious operations instead of attempting to prevent and then mitigate spurious operations. Entergy enhanced the procedure to address these concerns. NRC Inspection Report 2004-004 closed the original URI and opened a new item, URI 050000382/2004004-05 to document that it was impractical to follow up the original concern due to the procedure revision, and to document a plan to inspect the new revision.

On December 12 and 13, 2005, NRC inspectors observed timed walkthroughs of Revision 9 of the procedure. These walkthroughs confirmed completion of the procedure revision and verified that the licensee could meet the critical times determined by the Safe Shutdown

Analysis.

This URI is closed.

4OA6 Meetings, Including Exit

.1 On December 16, 2005, the inspector presented the inspection results to

Mr. J. Venable, Vice President, Operations, and other members of the staff, who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

.2 On August 18, 2005, the inspectors briefed Mr. R. Fletcher, Manager, Training and

Development, and other members of the licensee's management of the results of the inspection. The licensee acknowledged the findings presented. After final review of the overall biennial requalification written exams the inspectors conducted a teleconference exit with the licensee on December 6, 2005.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.3 On December 16, 2005, the inspectors briefed Mr. J. Venable and other members of the

licensee's management at the conclusion of the inspection. The licensee acknowledged the findings presented.

The inspectors returned all proprietary information to the licensee before the end of the inspection.

.4 On December 20, 2005, the inspector presented the inspection results to

Mr. J. Venable, Vice President, Operations, and other members of his staff who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

.5 On January 10, 2005, the resident inspectors presented the inspection results to

Mr. K. Walsh, General Plant Manager, and other members of licensee management at the conclusion of the inspection. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Anders, Superintendent, Plant Security

K. Boudreaux, Maintenance Rule Team Member

E. Brauner, Engineering Lead, Maintenance Rule

J. Brawley, ALARA Coordinator, Radiation Protection

B. Collyer, Senior Lead Engineer, Fire Protection

L. Dauzat, Supervisor, Radiation Protection Operations

C. Fugate, Assistant Manager, Operations (Shift)

T. Gaudet, Manager, Quality Assurance

J. Hall, Supervisor, Operations Training

A. Harris, Acting Director, Nuclear Safety Assurance

R. Jones, Supervisor, Simulator Support

T. Jones, Lead Licensed Operator, Requalification Backup

S. Landry, Specialist, Radwaste, Radiation Protection

J. Laque, Acting Director, Engineering

J. Lewis, Manager, Emergency Preparedness

B. Morrison, Coordinator, Maintenance Rule

R. Murillo, Acting Manager, Licensing

D. Newman, Supervisor, Radiation Protection

R Peters, Director, Planning and Scheduling

C. Pickering, Code Programs Engineer

B. Pilutti, Manager, Radiation Protection

O. Pipkins, Senior Licensing Engineer

G. Scott, Licensing Engineer

J. Venable, Vice President, Operations

K. Walsh, General Manager, Plant Operations

P. Wood, Reactor Operator

Attachment

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000382/2005005-01 NCV Inadequate Reactor Coolant System Leakage Detection System

Closed

05000382/2005004-00 LER Loss of Offsite Power During Hurricane Katrina

05000382/2004004-05 URI Review Revised Procedure and Re-perform Control Room Evacuation Walkthrough

05000382/2005005-01 NCV Inadequate Reactor Coolant System Leakage Detection System Attachment