IR 05000458/2008003

IR 05000458-08-003, on 03/30/2008 - 06/28/2008, River Bend Station; Maintenance Risk Assessments and Emergency Work Control, and Access Control to Radiologically Significant Areas

ML082110291
Person / Time
Site:	River Bend
Issue date:	07/29/2008
From:	Geoffrey Miller NRC/RGN-IV/DRP/RPB-C
To:	Mike Perito Entergy Operations
References
IR-08-003
Download: ML082110291 (40)

Text

July 29, 2008

Michael Perito Vice President, Operations Entergy Operations, Inc.

River Bend Station 5485 US Highway 61N St. Francisville, LA 70775

Subject: RIVERBEND STATION - NRC INTEGRATED INSPECTION REPORT 05000458/2008003

Dear Mr. Perito:

On June 28, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your River Bend Station. The enclosed report documents the inspection results, which were discussed on July 8, 2008, with you and other member of your staff.

The 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.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

The report documents three NRC-identified and self-revealing findings of very low safety significance. All three of these findings were determined to involve violations of NRC requirements. Additionally, four licensee-identified violations, which were determined to be of very low safety significance, are listed in this report. However, because of their very low safety significance, and because they have been entered into your corrective action program, the NRC is treating these findings as noncited violations, consistent with Section VI.A.1 of the NRC Enforcement Policy.

If you contest the subject or severity of any noncited violation in this report, 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 a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd., Suite 400, Arlington, Texas 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector at River Bend Station.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

UNITED STATES NUCLEAR REGULATORY COMMISSION R E GI ON I V 612 EAST LAMAR BLVD, SUITE 400 ARLINGTON, TEXAS 76011-4125

Entergy Operations, Inc.

- 2 -

Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

/RA/

Geoffrey B. Miller, Chief

Project Branch C

Division of Reactor Projects

Docket: 50-458 License: NPF-47

Enclosure:

NRC Inspection Report 05000458/2008003 w/Attachment: Supplemental Information

REGION IV==

Docket:

50-458

License:

NPF-47

Report:

05000458/2008003

Licensee:

Entergy Operations, Inc.

Facility:

River Bend Station

Location:

5485 U.S. Highway 61 St. Francisville, LA

Dates:

March 30 through June 28, 2008

Inspectors:

G. Larkin, Senior Resident Inspector, Project Branch C

C. Norton, Resident Inspector, Project Branch C

P. Goldberg, PE, Reactor Inspector, Engineering Branch 2

B. Baca, Health Physicist, Plant Support Branch

P. Elkmann, Sr. Emergency Preparedness Inspector, Operations Branch

B. Tindell, Resident Inspector, Project Branch A

Approved By:

Geoffrey B. Miller, Chief Project Branch C

Division of Reactor Projects

Enclosure-2-

SUMMARY OF FINDINGS

IR 05000458/2008003; 03/30/2008 - 06/28/2008; River Bend Station, Maintenance Risk

Assessments and Emergent Work Control, and Access Control to Radiologically Significant Areas

The report covered a 3-month period of routine baseline inspections by resident inspectors and announced baseline inspections by regional specialist inspectors. Three Green noncited violations were identified. The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the significance determination process 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 4, dated December 2006.

Cornerstone: Initiating Events

Green.

The inspectors identified a noncited violation of 10 CFR 50.65(a)(4)when operators failed to perform an adequate risk assessment associated with a reactor start-up while performing troubleshooting, and during maintenance activities on the main turbine electro hydraulic control system. This resulted in unanticipated oscillations in reactor power and pressure. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-4284.

Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), this finding is more than minor because Entergys risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. The risk assessment also failed to consider emergent maintenance activities that could increase the likelihood of initiating events. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component of decision making because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action H.1(b) (Section 1R13).

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR 50.65(a)(4)when operators failed to perform an adequate risk assessment while the Division 1 control building chilled water and control building air conditioning systems were unavailable. Specifically, the inspectors identified that licensee personnel nonconservatively evaluated the on-line risk as Green instead of

Yellow.

This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-2687.

Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected period is less than 1.0E-6. This finding has a crosscutting aspect in the area of problem identification and resolution component of operating experience because Entergy did not systematically communicate to affected internal stakeholders in a timely manner relevant internal operating experience P.2(a) (Section 1R13).

Cornerstone: Occupational Radiation Safety

Green.

The inspector reviewed a self-revealing noncited violation of Technical Specification 5.4.1 which resulted from workers failing to follow a radiation protection procedure. On January 13, 2008, three workers attempted to exit the radiologically controlled area and alarmed the personnel contamination monitors.

The workers were removing tubes from the Water Box B. The licensee determined radiation protection staff did not follow the radiation work permit planning procedure to use representative radiological surveys for the work performed. The radiation work permit planning did not include previous water box internal and other related surveys which would correspond to the removal of the water box tubes. The licensees investigation found that the contamination levels on the tubes were as high as 150,000 disintegrations per minute per 100 cm2. The licensee revised the radiation work permit to include the actual working conditions and appropriate personnel protective equipment.

Workers failing to follow a radiation protection procedure is a performance deficiency. The finding is greater than minor because, if left uncorrected, the deficiency would become a more radiologically significant safety concern resulting in additional workers unplanned, unintended dose as work continued to be performed under an inadequate radiation work permit. Since this issue involved workers unplanned and unintended dose, the Occupational Radiation Safety Significance Determination Process was used to determine the safety significance. The inspector determined the finding had very low safety significance because: (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, the finding had a crosscutting aspect in the area of human performance, work control, because the radiation protection staff did not plan work activities consistent with radiological safety by incorporating risk insights and job site conditions of the actual work to be performed during the radiation work permit planning [.3(a)]

(Section 2OS1).

Licensee-Identified Violations

Violations of very low safety significance that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

River Bend Station began the inspection period at 100 percent reactor power. The plant remained at 100 percent power except for short periods to adjust the existing rod pattern. The rod pattern adjustment resulted in reducing reactor power on April 5, 2008, to 86 percent power, May 10 to 55 percent power, May 13 to 86 percent power, and June 6 to 87 percent power.

Entergy made one additional down power on June 27 to 13 percent power to enter the drywell and add oil to reactor recirculation Pump A lower motor oil reservoir.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

Summer Readiness for Offsite and Alternate AC Power

a. Inspection Scope

The inspectors reviewed River Bend Station plant procedures and interviewed operations shift management to verify that plant features and procedures for operation and continued availability of offsite and alternate AC power systems during adverse weather were appropriate. The inspectors also reviewed portions of the Corporate Entergy Nuclear Management Manual and interviewed licensing personnel to verify the transmission system operators (TSO) procedures for operation and continued availability of offsite and alternate AC power systems during adverse weather were appropriate.

The inspectors reviewed the licensees nuclear power plant (NPP) procedures and the Entergy Nuclear Management Manual affecting these areas and the communication protocols between the TSO and the NPP to verify that the appropriate information was exchanged when issues arose that could impact the offsite power system. Some examples of information to be conveyed include: coordination between the TSO and the NPP during an off-normal or emergency event affecting the NPP, explanation of the event, an estimate of when the offsite power system will be returned to a normal state, and notification to the NPP when the offsite power system was returned to normal. The inspectors verified that these NPP and TSO procedures address measures to monitor and maintain availability and reliability of both the offsite AC power system and the onsite alternate AC power system prior to and during adverse weather conditions.

Specifically, these procedures addressed:

• The actions to be taken when notified by the TSO that the posttrip voltage of the offsite power system at the NPP will not be acceptable to assure the continued operation of the safety-related loads without transferring to the onsite power supply

• The compensatory actions identified to be performed if it is not possible to predict the posttrip voltages at the NPP for the current conditions

• Required re-assessment of plant risk based on maintenance activities which could affect grid reliability, or the ability of the transmission system to provide offsite power

• Required communications between the NPP and the TSO at the NPP could impact the transmission system, or when the capability of the transmission to provide adequate offsite power is challenged, even though the TSO is independent of the NPP license

Documents reviewed by the inspectors are listed in the attachment.

This inspection constitutes one sample for summer weather effects on offsite and alternate AC power as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

Quarterly Partial System Walkdowns (71111.04Q)

a. Inspection Scope

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

• Instrument air system

• Division 2 125-Volt DC power system (ENB)

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 impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report (USAR), Technical Specification (TS)requirements, Administrative TS, outstanding work orders (WOs), condition reports (CRs), 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 walked down 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 (CAP) with the appropriate significance characterization.

These activities constituted two partial system walkdown samples as defined by Inspection Procedure 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

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:

• May 29, 2008, Reactor Building, Fire Area RC-3/Z-1, 2, 3, and 4

• May 29, 2008, Reactor Building, Fire Area RC-4/Z-1, 2, 3, and 4

• June 4, 2008, Control Building, Division 1 Standby Switchgear Room and Cable Case, 98-foot level, Fire Area C-15 and C-10

• June 4, 2008, Control Building, Division 2 Standby Switchgear Room and Cable Case, 98-foot level, Fire Area C-14 and C-2

• June 4, 2008, Control Building, Division 1, Water Chiller Equipment Room and Cable Case C-9, 98-foot level, Fire Area C-13E and C-1

• June 4, 2008, Control Building, Division 2, Water Chiller Equipment Room and Cable Case C-9, 98-foot level, Fire Area C-13W and C-9

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 licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants 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 plants 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 licensees CAP. Documents reviewed by the inspectors are listed in the attachment.

These activities constituted six quarterly fire protection inspection samples as defined by Inspection Procedure 71111.05-05.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

Internal Flooding

a. Inspection Scope

The inspectors:

(1) reviewed the USAR, the flooding analysis, and plant procedures to assess seasonal susceptibilities involving internal flooding;

(2) reviewed the USAR and CAP to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of

(a) sump pumps,

(b) level alarm circuits,

(c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes;

(4) verified that operator actions for coping with flooding can reasonably achieve the desired outcomes; and

(5) walked down the area listed below to verify the adequacy of:

(a) equipment seals located below the floodline,

(b) floor and wall penetration seals,

(c) watertight door seals,

(d) common drain lines and sumps,

(e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

• May 16, 2008, Unit 1, Tunnel B, emergency core cooling system cubicle doors and sump pumps

Documents reviewed by the inspectors are listed in the attachment.

These activities constituted one flood protection measures inspection sample as defined by Inspection Procedure 71111.06-05.

b. Findings

No findings of significance were identified.

1R07 Triennial Heat Sink Performance

.1 Performance Testing, Maintenance, and Inspection Activities

a. Inspection Scope

Inspection Module 71111.07, Heat Sink Performance, requires that two to three safety-related heat exchangers, either directly or indirectly connected to the safety-related service water system, be reviewed to ensure they are either tested or inspected and cleaned. The inspector selected the following three heat exchangers that were ranked high in the plant specific risk assessment and are connected to the safety-related service water system:

• Emergency diesel generator jacket water heat exchanger

• Residual heat removal heat exchanger

• The standby cooling water tower, which is the ultimate heat sink

For the heat exchangers directly connected to the safety-related service water system, the inspector reviewed whether testing, or inspection and cleaning, and maintenance, and the fouling monitoring program provided sufficient controls to ensure proper heat transfer. The inspector reviewed chemical controls used to avoid fouling, heat exchanger testing results, and inspection and cleaning results. The inspector walked down the three samples and reviewed the method used to ensure that the ultimate heat sink had sufficient water at all times. In addition, the inspector reviewed other sources of water that could be used to fill the heat sink during emergency conditions.

For the chosen heat exchangers, the inspector verified the proper extrapolation of test conditions to design conditions, appropriate use of test instrumentation, and appropriate accounting for instrument inaccuracies. The inspector reviewed the methods and results of heat exchanger inspection and cleaning, verified that the methods used to inspect and clean were consistent with industry standards, and ensured that the as-found results were appropriately dispositioned such that the final conditions were acceptable. The inspector reviewed trending of the heat exchangers. Additionally, the inspector verified that the licensee appropriately trended these inspection and cleaning results, assessed the causes of the trends, and took necessary actions for any step changes in these trends.

These activities constituted three heat sink inspection samples as defined by Inspection Procedure 71111.07-05.

b. Findings

No findings of significance were identified.

.2 Verification of Conditions and Operations Consistent with Design Bases

a. Inspection Scope

For the selected heat exchangers, the inspector verified that the licensee established heat sink and heat exchanger conditions and operation that were consistent with the design assumptions. Specifically, the inspector reviewed the applicable calculations to ensure that the thermal performance test acceptance criteria for the heat exchangers were being applied consistently throughout the calculations. In addition, the inspector reviewed test data for the heat exchangers and design and vendor-supplied information to ensure that the heat exchangers were within their design bases.

b. Findings

No findings of significance were identified.

.3 Identification and Resolution of Problems

a. Inspection Scope

The inspector verified that the licensee had entered significant heat exchanger/heat sink performance problems into the CAP. The inspector reviewed approximately ten CRs, which are listed in the attachment.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On June 24, 2008, the inspectors observed operator team testing of senior reactor operators and reactor operators to identify deficiencies and discrepancies in the training, to assess operator performance, and to assess the evaluator's critique. The training scenario involved a failed turbine plant component cooling water pump, a failed turbine plant component cooling water valve actuation, a stuck open safety relief valve, entry into the emergency operating procedures, and emergency classification.

Documents reviewed by the inspectors included:

• Simulator Scenario RSMS-OPS-431, Stuck Open SRV, Revision 6

This inspection constitutes one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee'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:

• Division I diesel generator postsurveillance air roll on May 8, 2008

• Main turbine speed sensing erratic speed indication on March 7, 2008

• Division 2 control building chilled water system outage on April 7-9, 2008

• Station blackout diesel ground fault unavailability on May 9, 2008 The inspectors selected these activities based on their 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 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 TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Documents reviewed by the inspectors are listed in the attachment.

These activities constituted four maintenance risk assessments and emergent work control inspection samples as defined by Inspection Procedure 71111.13-05

b. Findings

===.1

=

The inspectors identified a Green noncited violation (NCV) of 10 CFR 50.65(a)(4) involving the failure of operators to perform an adequate risk assessment while the Division 1 control building chilled water and control building air conditioning systems were unavailable. Specifically, the inspectors identified that licensee personnel nonconservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis (PSA) evaluation.

Description.

On April 9, 2008, the licensee removed the Division 1 control building air conditioning (HVC), air handling Unit 1A and Division I control building chilled water (HVK) Chillers A and C from service for scheduled maintenance. Entergy assessed the plant risk as Green with a plant safety index of 10.0. Because recent PRA assumptions increased the significance of these systems to plant safety, the actual risk was Yellow with a plant safety index of 9.5. Entergy failed to perform an adequate risk assessment because instructions from Entergys PSA group to model risk when HVK and HVC systems are unavailable were not appropriately implemented. The PSA group originally distributed interim measures via email that was not formally included in procedures or instructions provided to station operators. As a result of this finding, Entergy issued a standing order to implement the PSA group interim measures. As a final corrective measure, the PSA group will update the on-line risk calculation program to include actual risk associated with the removal of HVK and HVC components.

Analysis.

The inspectors determined that the licensees failure to perform an adequate risk assessment was a performance deficiency. Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of problem identification and resolution component of operating experience because Entergy did not systematically communicate to affected internal stakeholders in a timely manner relevant internal operating experience P.2(a).

Enforcement.

10 CFR 50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, requires, in part, that, prior to performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee failed to perform an adequate risk assessment before performing maintenance on the control building chilled water and control building air condition systems from April 7-10, 2008. Because the finding was of very low safety significance and has been entered into the licensees CAP as Condition Report RBS-2008-02687, this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000458/2007003-01, Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service.

===.2

=

The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) involving the failure of operators to perform an adequate risk assessment associated with a reactor start-up while performing troubleshooting and during maintenance activities on the main turbine electro hydraulic control (EHC) system. This resulted in unanticipated oscillations in reactor power and pressure.

Description.

On March 5, 2007, the reactor scrammed on reactor high pressure. The cause of the reactor high pressure was traced to an unexpected closure of the turbine control valves in response to a faulty turbine speed control signal. With the reactor in cold shutdown, Entergy began troubleshooting the EHC system to determine the cause of the faulty speed control signal. Utilizing Procedure EN-MA-125, Troubleshooting and Control of Maintenance Activities, and WO 1141966, Main Turbine Trip, the risk assessment for this troubleshooting activity was appropriately classified Level 3, equipment removed from service or troubleshooting activities do not affect the operation or safety of the plant. While still troubleshooting EHC speed control logic circuitry, Entergy commenced a reactor startup and placed EHC in pressure control on the bypass valves. Plant operators failed to update the risk assessment for this new condition. Troubleshooting activities continued and maintenance technicians determined that a logic card needed replacement in the speed control circuitry. In order to isolate speed control logic from pressure control logic, the technicians removed an amplifier card that feeds a signal from the speed control logic to the pressure control logic.

Entergy did not assess the risk associated with this proposed maintenance activity nor did they develop risk management actions for handling consequences of any increased risk associated with this action. When the amplifier card was subsequently removed, the bypass valves closed. Reactor pressure and power increased causing operators to take the unplanned and unanticipated actions of opening steam line drains to control reactor pressure and up ranging several IRM channels in order to maintain adequate scram set point margin.

The inspectors concluded that Entergy had opportunities to screen the work scope as Risk Level 1 for equipment that is in service and presents a risk of tripping the plant or causing a plant transient. Risk Level 1 work requires additional work controls such as, work approval by the operations manager, a detailed work plan, direct supervision, system engineer involvement and, readiness reviews to minimize all possible mechanical and electrical risks or plant challenges. With an appropriate risk analysis and the subsequent additional work controls the above described loss of reactivity control may have been prevented.

Analysis.

The inspectors determined that failure to update the risk assessment prior to starting the reactor and placing EHC in pressure control while performing troubleshooting activities on EHC was a performance deficiency. Also failure to perform a risk assessment prior to performing maintenance activities on EHC while controlling reactor with EHC was a performance deficiency. These performance deficiencies resulted in unanticipated oscillations in reactor power and pressure. Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), this finding is more than minor because the risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. The risk assessment also failed to consider emergent maintenance activities that could increase the likelihood of initiating events. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component of decision making because the licensee did not use conservative assumptions in decision making and adopts a requirement to demonstrate the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1.(b)].

Enforcement.

10 CFR 50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, requires, in part, that, prior to performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, Entergy failed to re-assess risk associated with commencing a reactor start up while performing troubleshooting activities on the EHC system logic. Also contrary to the above, Entergy failed to perform a risk assessment prior to performing maintenance activities on EHC logic while controlling reactor pressure with EHC. This resulted in unanticipated oscillations in reactor power and pressure requiring the operators to up range several IRM channels in order to maintain adequate reactor scram set point margin. Because the finding was of very low safety significance and has been entered into the licensees CAP as Condition Report CR-RBS-2008-04284, this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 0500458/2008003-02; Inadequate Risk Assessment While Troubleshooting Results in Unanticipated Reactor Power Oscillations.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• CR-RBS-2008-03202, Division I diesel generator intake manifold knee brace loose bolts following a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance run, reviewed on May 9, 2008

• CR-RBS-2008-02163, erratic turbine bypass valve movement during EHC system control valve amplifier card replacement, reviewed on May 9, 2008

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 TS 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 TS and USAR to the licensees 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. Documents reviewed are listed in the attachment.

This inspection constitutes two operability evaluations inspection samples as defined in Inspection Procedure 71111.15-05.

b. Findings

No findings of significance 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:

• WO 00141966, Tasks 3 and 4, Replace Speed Sensors 4 and 5, reviewed on May 1, 2008

• WO 00141966, Tasks 8 and 9, I & C H13-P821 Replace Turbine EHC Load Gate Amp Card 1L1-M001, reviewed on June 12, 2008

• WO 00149354, BYS-EG1 - Test, Verify, Station Blackout Portable DG, reviewed on June 12, 2008

The inspectors selected these activities based upon the structures, systems, or components (SSCs) ability to affect risk. The inspectors evaluated these activities for the following (as applicable):

(1) the effect of testing on the plant had been adequately addressed,

(2) testing was adequate for the maintenance performed,

(3) acceptance criteria were clear and demonstrated operational readiness,

(4) test instrumentation was appropriate,

(5) tests were performed as written in accordance with properly reviewed and approved procedures,

(6) equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion), and

(7) test documentation was properly evaluated. The inspectors evaluated the activities against TS, the USAR, 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 into the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the attachment.

This inspection constitutes three postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the USAR, procedure requirements, and TS to ensure that the surveillance activity listed below demonstrated that the SSCs tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate:

(1) preconditioning;

(2) evaluation of testing impact on the plant;

(3) acceptance criteria;

(4) test equipment;

(5) procedures;

(6) jumper/lifted lead controls;

(7) test data;

(8) testing frequency and method demonstrated TS operability;

(9) test equipment removal;

(10) restoration of plant systems;

(11) fulfillment of ASME Code requirements;

(12) updating of performance indicator (PI) data;

(13) engineering evaluations, root causes, and bases for returning tested SSCs not meeting the test acceptance criteria were correct;

(14) reference setting data; and

(15) annunciators and alarms setpoints.

The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing.

• SOP-0054, Station Blackout Diesel Generator (Monthly Run), Revision 301, observed and reviewed on May 9, 2008

• STP-309-0603, Division III ECCS Test, Revision 27, test performed on April 16, 2008

• STP-309-0613, Division III Diesel Generator 24 Hour Run, Revision 15, performed on April 17, 2008

Documents reviewed by the inspectors are listed in the attachment.

This inspection constitutes three surveillance testing inspection samples as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings of significance were identified.

1EP1 Exercise Evaluation

a. Inspection Scope

The inspectors reviewed the objectives and scenario for the 2008 Biennial Emergency Preparedness Exercise to determine if the exercise would acceptably test major elements of the emergency plan. The scenario simulated a dropped control rod causing fuel failure, an electrical grid perturbation that tripped reactor feedwater pumps, and a pipe failure inside the reactor drywell, resulting in reactor water level lowering to below the top of active fuel. Multiple fission product barrier failures followed, including core damage and the failure of a drywell airlock door seal resulting in a filtered and monitored radiological release to the environment via the standby gas treatment system. The scenario was designed to demonstrate the licensee's capabilities to implement the emergency plan.

The inspectors evaluated exercise performance by focusing on the risk-significant activities of event classification, offsite notification, recognition of offsite dose consequences, and development of protective action recommendations, in the control room simulator and the following dedicated emergency response facilities:

$

Technical Support Center

$

Operations Support Center

$

Emergency Operations Facility

The inspectors also assessed recognition of and response to abnormal and emergency plant conditions, the transfer of decision making authority and emergency function responsibilities between facilities, onsite and offsite communications, protection of emergency workers, emergency repair evaluation and capability, and the overall implementation of the emergency plan to protect public health and safety and the environment. The inspectors reviewed the current revision of the facility emergency plan, and emergency plan implementing procedures associated with operation of the above facilities and performance of the associated emergency functions. These procedures are listed in the attachment to this report.

The inspectors compared the observed exercise performance with the requirements in the facility emergency plan; 10 CFR 50.47(b); 10 CFR Part 50, Appendix E; and with the guidance in the emergency plan implementing procedures and other federal guidance.

The inspectors attended the post-exercise critiques in each emergency response facility to evaluate the licensees initial self-assessment of exercise performance. The inspectors also attended a subsequent formal presentation of critique items to plant management.

This inspection constitutes one sample as defined in Inspection Procedure 71114.01-05.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed an in-office review of Revision 32 to the River Bend Station Emergency Plan. This revision clarified the emergency response responsibilities of the shift technical advisor, updated accidents analyzed in the Updated Safety Analysis Report using the NRC-approved alternate source term, revised the means to display radiation monitor and meteorological information in licensee emergency response facilities from the RM-21 system to the emergency response information system, revised the predesignated offsite dose calculation manual sampling locations, incorporated the Year 2000 census data, updated the titles of Louisiana state agencies, and made minor editorial and administrative corrections.

The revision was compared to its previous revision, to the criteria of NUREG-0654, ACriteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,@ Revision 1, and to the standards in 10 CFR 50.47(b) to determine if the revision adequately implemented the requirements of 10 CFR 50.54(q). This review was not documented in a safety evaluation report and did not constitute approval of the licensees changes, therefore, the revision is subject to future inspection.

This inspection constitutes one sample as defined in Inspection Procedure 71114.04-05.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control To Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the Technical Specifications, and the licensees procedures required by Technical Specifications as criteria for determining compliance.

During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:

• PI events and associated documentation packages reported by the licensee in the Occupational Radiation Safety Cornerstone

• Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms

• Adequacy of the licensees internal dose assessment for any actual internal exposure greater than 50 millirem committed effective dose equivalent

• Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individual deficiencies

• Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas

The inspector completed 9 of the required 21 samples.

b. Findings

Introduction.

The inspector reviewed a self-revealing NCV of Technical Specification 5.4.1 which resulted from workers failing to follow a radiation protection procedure.

Description.

On January 13, 2008, three workers attempted to exit the radiologically controlled area and were determined to be contaminated. These workers were removing tubes from the Water Box B. The licensee investigated the event and determined that:

(1) the workers did not use the radiation work permits prescribed engineering controls for the work activity, and

(2) the surveys used in the radiation work permit planning were not representative of the actual work being performed. However, when the inspector reviewed the event and the licensees apparent cause evaluation, the inspector determined the licensees evaluation was in error and an incorrect radiation work permit revision was used. The radiation work permit in effect during the work activity did not contain engineering controls as stated in the licensees apparent cause evaluation. The planning for the radiation work permit used surveys which indicated the work would be conducted in a noncontaminated area and were not representative for the removal of the water box tubes (previous water box internal and other related system surveys). The investigation found that tube contamination levels were as high as 150,000 disintegrations per minute per 100 cm2.

Despite the use of an incorrect radiation work permit revision in the apparent cause evaluation, appropriate corrective actions were taken in response to this event. The licensee revised the radiation work permit to include the actual working conditions and appropriate personnel protective equipment. The licensee wrote Condition Report RBS-2008-3517 to address the apparent cause evaluation error.

This finding was self-revealing because the event was brought to the licensees attention when the workers alarmed the personnel contamination monitors.

Analysis.

Workers failing to follow a radiation protection procedure is a performance deficiency. The finding is greater than minor because, if left uncorrected, the deficiency would become a more radiologically significant safety concern resulting in additional workers unplanned, unintended dose. Since this issue involved workers unplanned and unintended dose, the Occupational Radiation Safety Significance Determination Process was used to determine the safety significance. The inspector determined the finding had very low safety significance because:

(1) it was not an ALARA finding,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

Additionally, the finding had a crosscutting aspect in the area of human performance, Work Control, because the radiation protection staff did not plan work activities consistent with radiological safety by incorporating risk insights and job site conditions of the actual work to be performed during the radiation work permit planning H.3(a).

Enforcement.

Technical Specification 5.4.1 requires procedures in accordance with Regulatory Guide 1.33, Appendix A. Section 7(e) of the Regulatory Guide requires procedures for access control to radiation areas including a radiation work permit system. Procedure EN-RP-105, Radiation Work Permits, Revision 4, Section 5.3, Step [5] states, in part, radiation protection/ALARA will review historical or current radiological information for the area and the work to be performed. Radiation protection staff violated this requirement when they failed to review historical or current radiological information and use the most representative surveys for the actual work to be performed during the planning and writing of the associated radiation work permit. Because this violation was of very low safety significance and was entered into the licensees CAP as Condition Report RBS-2008-0340, it is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000458/2008003-03, Failure to Follow a Radiation Protection Procedure.

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

• Ten work activities from previous work history data and during the last outage which resulted in the highest personnel collective exposures

• Site-specific trends in collective exposures, plant historical data, and source-term measurements

• Site-specific ALARA procedures

• ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements

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

• Integration of ALARA requirements into work procedure and radiation work permit documents

• Person-hour estimates provided by maintenance planning and other groups to the radiation protection group with the actual work activity time requirements

• Shielding requests and dose/benefit analyses

• Dose rate reduction activities in work planning

• Postjob (work activity) reviews

• Assumptions and basis for the current annual collective exposure estimate, the methodology for estimating work activity exposures, the intended dose outcome, and the accuracy of dose rate and man-hour estimates

• Method for adjusting exposure estimates, or replanning work, when unexpected changes in scope or emergent work were encountered

• Exposure tracking system

• Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding

• Exposures of individuals from selected work groups

• Specific sources identified by the licensee for exposure reduction actions, priorities established for these actions, and results achieved since the last refueling cycle

• Declared pregnant workers during the current assessment period, monitoring controls, and the exposure results

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

• Resolution through the corrective action process of problems identified through post-job reviews and post-outage ALARA report critiques

• Corrective action documents related to the ALARA program and follow-up activities, such as initial problem identification, characterization, and tracking

• Effectiveness of self-assessment activities with respect to identifying and addressing repetitive deficiencies or significant individual deficiencies

The inspector completed 13 of the required 15 samples and 10 of the optional samples.

b. Findings

No findings of significance were identified.

4OA1 PI Verification

Cornerstone: Initiating Events

a. Inspection Scope

The inspectors sampled licensee submittals for the three PIs listed below for the period January 1 to March 29, 2008. The definitions and guidance of NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 5, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period. The inspectors reviewed licensee event reports, monthly operating reports, and operating logs as part of the assessment.

• Unplanned Scrams Per 7,000 Critical Hours

• Unplanned Power Changes per 7,000 Critical Hours

• Unplanned Scrams with Complications

The inspectors completed three inspection samples.

b. Findings

No findings of significance were identified.

Cornerstone: Barrier Integrity

a. Inspection Scope

The inspectors sampled licensee submittals for the two PIs listed below for the period January 1, 2007, to March 31, 2008. The definitions and guidance of NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 5, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period. The inspectors:

(1) reviewed reactor coolant system (RCS) chemistry sample analyses for dose equivalent Iodine-131 and compared the results to the TS limit;

(2) observed a chemistry technician obtain and analyze an RCS sample;

(3) reviewed operating logs and surveillance results for measurements of RCS identified leakage; and

(4) observed a surveillance test that determined RCS identified leakage.

• RCS Specific Activity

• RCS Leakage

The inspectors completed two inspection samples.

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

a. Inspection Scope

Drill and Exercise Performance

Emergency Response Organization Participation

Alert and Notification System Reliability

The inspectors reviewed licensee evaluations for the three emergency preparedness cornerstone PIs for the period April 2007 through March 2008. The definitions and guidance of NEI 99-02, ARegulatory Assessment Indicator Guideline,@ Revisions 3 and 4; the licensees PI Procedures EN-LI-114, APerformance Indicator Process,@ Revisions 2 and 3, and EPP-2-701, Prompt Notification System Maintenance and Testing, Revision 20, were used to verify the accuracy of the licensee=s evaluations for each PI reported during the assessment period.

The inspectors reviewed a 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 inspectors reviewed 35 selected emergency responder qualification, training, and drill participation records. The inspectors reviewed alert and notification system testing procedures, maintenance records, and a sample of siren test records.

The inspectors also reviewed other documents listed in the attachment to this report.

The inspectors completed three samples.

b. Findings

No findings of significance were identified.

Cornerstone: Occupational Radiation Safety

a. Inspection Scope

Occupational Exposure Control Effectiveness

The inspector reviewed licensee documents from April 1, 2007, through March 31, 2008.

The review included corrective action documentation that identified occurrences in locked high radiation areas (as defined in the licensees Technical Specifications), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02, "Regulatory Assessment Indicator Guideline,"

Revision 5). Additional records reviewed included ALARA records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the PI data. In addition, the inspector toured plant areas to verify that high radiation, locked high radiation, and very high radiation areas were properly controlled. PI definitions and guidance contained in NEI 99-02, Revision 5, were used to verify the basis in reporting for each data element.

The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety

a. Inspection Scope

Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences

The inspector reviewed licensee documents from April 1, 2007, through March 31, 2008.

Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded PI thresholds and those reported to the NRC. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the PI data. PI definitions and guidance contained in NEI 99-02, Revision 5, were used to verify the basis in reporting for each data element.

The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into the licensees CAP.

This assessment was accomplished by reviewing WOs, condition reports, and attending corrective action review and work control meetings. The inspectors:

(1) verified that equipment, human performance, and program issues were being identified by the licensee at an appropriate threshold and that the issues were entered into the CAP;

(2) verified that corrective actions were commensurate with the significance of the issue;

and

(3) identified conditions that might warrant additional follow-up through other baseline inspection procedures.

b. Findings

No findings of significance were identified.

.2 Semiannual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees CAP and associated documents to identify trends associated with mechanical degradation of emergency diesel generators such as loose bolting, subcomponent cracking and fluid system leakage 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 4OA2.1 above, licensee trending efforts, and licensee human performance results. The inspectors nominally considered the 6-month period of December 1, 2007, through May 30, 2008, although some examples expanded beyond those dates where the scope of the trend warranted.

The inspectors also included issues documented outside the normal CAP 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 licensees CAP trending reports. Corrective actions associated with a sample of the issues identified in the licensees trending reports were reviewed for adequacy.

This review constituted a single semi-annual trend inspection sample.

b. Findings

No findings of significance were identified.

.3 Annual Sample Review

a. Inspection Scope

The inspectors reviewed licensee evaluation reports for drills conducted between June 2006 and May 2008, and reviewed a summary list of issues associated with emergency response organization performance. The drill evaluation reports were reviewed to identify emergency response organization weaknesses and deficiencies, and the summary of condition reports was reviewed to identify the range of emergency response organization and emergency response facility performance issues.

The inspectors observed the June 11, 2008, biennial emergency preparedness exercise to verify that previously-identified emergency response organization weaknesses and deficiencies, and emergency response facility problems, had been corrected.

b. Findings

No findings of significance were identified.

.4 Radiation Safety

The inspector evaluated the effectiveness of the licensees problem identification and resolution process with respect to the following inspection areas:

• Access Control to Radiologically Significant Areas (Section 2OS1)

• ALARA Planning and Controls (Section 2OS2)

b. Findings

Section 2OS1 describes a finding in which the inspector identified that an incorrect radiation work permit was used during a licensees apparent cause evaluation.

Section 4OA7 documents a licensee identified finding that was not captured in the licensees CAP at the time of the event. The inspector reviewed the licensees documentation for internal contaminations and internal dose assignments that occurred during the January through February 2008 refueling outage. The inspector found three internal contamination events (to include the finding in Section 4OA7) that were not entered into the licensees CAP. The inspector did not identify any required corrective actions which were not performed for these three events. The licensee wrote Condition Report RBS-2008-3191 to capture this issue.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors performed the following observations of security force personnel and activities to ensure that the activities were consistent with licensees security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant work hours.

These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.

b. Findings

No findings of insignificance were identified.

4OA6 Meetings, Including Exit

Exit Meetings

On April 4, 2008, the inspector presented the results of the triennial heat sink performance inspection to Mr. M. Perito and other members of his staff who acknowledged the findings. The inspector did review proprietary information and agreed to destroy the information prior to the report being issued.

On May 30, 2008, the inspector presented the occupational radiation safety inspection results to Mr. M. Perito and other members of his staff who acknowledged the findings.

The inspector confirmed that proprietary information was not provided or examined during the inspection. The inspector re-exited on June 18, 2008, with licensing staff and discussed changes to two finding characterizations.

On June 25, 2008, the inspectors participated by teleconference in the licensees final Exercise Evaluation report to management and discussion of emergency response organization performance.

On June 26, 2008, the inspectors conducted a telephonic exit meeting to present the results of the 2008 Biennial Emergency Preparedness Exercise inspection to Mr. M. Perito, Vice President, Operations, and other members of his staff, who acknowledged the findings. The inspectors confirmed that proprietary, sensitive, or personal information examined during the inspection had been returned to their identified custodians.

On July 8, 2008, the inspectors presented the integrated baseline inspection results to Mr. M. Perito and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

4OA7 Licensee-Identified Violations

The following findings of very low safety significance were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as NCVs.

• Technical Specification 5.7.2 requires, in part, that areas with radiation levels greater than and equal to 1,000 millirem per hour shall be provided with locked or continuously guarded doors to prevent unauthorized access. Contrary to the Technical Specification requirement, the licensee discovered the drywell 95-foot elevation access was left unattended for approximately 5 minutes. The locked high radiation door guard left his/her post before being properly relieved.

The licensee promptly regained access control and captured the event in Condition Report RBS-2008-00387. The inspector determined the finding had very low safety significance because:

(1) it was not an ALARA finding,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

• Technical Specification 5.7.1 requires, in part, each high radiation area in which the intensity of radiation is greater than 100 millirem per hour but less than 1,000 millirem per hour shall be barricaded and conspicuously posted as a high radiation area. Contrary to the Technical Specification requirement, the licensee identified a high radiation area that was not barricaded and conspicuously posted. On February 3, 2008, a chemical decontamination worker notified radiation protection of a higher than expected peak dose rate for the reactor water clean up Heat Exchanger Room RB 141. Radiation protection performed a follow up survey and found a hose with a hot spot reading 2500 millirem per hour on contact and 500 millirem per hour at 30 centimeters. The radiation protection technician immediately barricaded and posted the area as a high radiation area.

The licensee captured the event in Condition Report RBS-2008-00387. The inspector determined the finding had very low safety significance because:

(1) it was not an ALARA finding,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

• Technical Specification 5.4.1 requires, in part, procedures in accordance with Regulatory Guide 1.33, Revision 2, Appendix A. Section 7(e) of the appendix includes procedures for contamination control and monitoring personnel exposures. Procedure EN-RP-100, Radworker Expectations, Revision 1, Section 5.6 (h), requires, in part, all tasks in radiologically controlled areas be completed with ALARA concepts in mind and that unintended exposure is not acceptable. Contrary to the procedure requirement, on February 9, 2008, a Westinghouse chemist used excessive force to remove highly contaminated samples from a holder (banging the sample holder on a work surface) that resulted in him becoming externally and internally contaminated. Using excessive force to remove highly contaminated samples, which could contaminate the worker and the work area, was not keeping with ALARA principles. After the chemist forcibly removed the samples, the attending radiation protection technician stopped the work activity and had the individual process out of the radiologically controlled area. The individual was assigned an internal dose of 63.9 millirem. The individual received personnel discipline. This event was not initially captured in the licensee CAP and is documented in Section 4OA2 of this report. The inspector determined the finding had very low safety significance because:

(1) it was not an ALARA finding,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and (4)the ability to assess dose was not compromised.

• Technical Specification 5.4.1 requires, in part, procedures in accordance with Regulatory Guide 1.33, Revision 2, Appendix A, Section 7(e), of the appendix includes procedures for access control to radiation areas including a radiation work permit system. Procedure EN-RP-100, Radworker Expectations, Revision 1, Section 5.3, Radiologically Controlled Area Access, step 8, states, The RWP shall be read, understood, and obeyed. Radiation Work Permit 08-1935-01, Revision 0, requires radiation protection coverage at system breaches. Contrary to the radiation work permit, on February 8, 2008, two workers became involved in work to remove the containment spray line Valve CSL-V30 and failed to notice they reached a radiation work permit hold point before breaching the containment spray line. The workers received internal contaminations and were assigned internal doses of 48.3 millirem and 10.9 millirem. The licensees immediate corrective action was to restrict the workers radiologically controlled area access pending a review of the event. The licensee performed briefings of the event with radiation protection staff and all members of the maintenance shops. The licensee captured the event in Condition Report RBS-2008-1473. The inspector determined the finding had very low safety significance because:

(1) it was not an ALARA finding,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Baccus, Supervisor, Radiation Protection

D. Burnett, Acting Manager, Corrective Action

M. Chase, Manager, Training and Development

J. Clark, Assistant Operations Manager - Training

C. Forpahl, Manager, Engineering Programs & Components

B. Heath, Superintendent, Chemistry

K. Higginbotham, Acting Manager, Operations

K. Higginbotham, Assistant Operations Manager - Shift

B. Houston, Manager, Radiation Protection

K. Huffstatler, Technical Specialist, Licensing

A. James, Manager, Security

J. Leavines, Manager, Emergency Preparedness

C. Stout, Manager, Plant Maintenance

D. Lorfing, Manager, Licensing

D. Myers, Specialist, Radiation Protection

J. Wilson, Supervisor, Reactor Engineering

W. Mashburn, Manager, Design Engineering

B. Matherne, Manager, Planning and Scheduling/Outage

R. McAdams, Manager, System Engineering

J. McElwain, Manager, Human Resources

E. Olson, General Manager, Plant Operations

E. Roan, Manager, Outage

J. Roberts, Director, Nuclear Safety Assurance

K. Rockwood, Supervisor, Radiation Protection

J. Schlesinger, Supervisor, Engineering

T. Tankersley, Manager, Quality Assurance

J. Venable, Senior Site Vice President

D. Wiles, Director, Engineering

NRC Personnel

E. Reusch, Reactor Inspector, DRS/EB2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000458/2008003-01 NCV Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service (1R13)

05000458/2008003-02 NCV Inadequate Risk Assessment While Troubleshooting Results in Unanticipated Reactor Power Oscillations (1R13)

05000458/2008003-03 NCV Failure to Follow a Radiation Protection Procedure

(Section 2OS1.2)

Closed

None

Discussed

None