IR 05000458/2010003

IR 05000458-10-003 on 04/01/2010 - 06/30/2010 for River Bend Station, Integrated Inspection Report; Maintenance Risk Assessments and Emergent Work Control;,Operability Evaluations; Occupational Dose Assessment

ML102170526
Person / Time
Site:	River Bend
Issue date:	08/05/2010
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-C
To:	Mike Perito Entergy Operations
References
IR-10-003
Download: ML102170526 (51)

Text

August 5, 2010

Michael Perito Site Vice President Entergy Operations, Inc. River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Subject: RIVER BEND STATION - NRC INTEGRATED INSPECTION REPORT 05000458/2010003

Dear Mr. Perito:

On June 30, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your River Bend Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 8, 2010, with Mr. B. Cox, Manager, Operations, and other members of your staff.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents one NRC-identified finding and two self-revealing findings of very low safety significance (Green). All of these findings were determined to involve violations of NRC requirements. Additionally, one licensee identified violation, which was determined to be of very low safety significance, is listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as noncited violations, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the violations or the significance of the noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the River Bend Station facility. In addition, if you disagree with the crosscutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at River Bend Station. Entergy Operations, Inc. - 2 -

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRC's document system (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Vincent G. Gaddy, Chief Project Branch C Division of Reactor Projects Docket: 50-458 License: NPF-47

Enclosure:

NRC Inspection Report 05000458/2010003

w/Attachment:

Supplemental Information cc w/

Enclosure:

Senior Vice President and COO Entergy Operations, Inc P. O. Box 31995 Jackson, MS 39286-1995 Vice President, Oversight Entergy Operations, Inc.

P. O. Box 31995 Jackson, MS 39286-1995 Senior Manager, Nuclear Safety & Licensing Entergy Nuclear Operations P. O. Box 31995 Jackson, MS 39286-1995 Manager, Licensing Entergy Operations, Inc. 5485 US Highway 61N St. Francisville, LA 70775 Entergy Operations, Inc. - 3 -

Attorney General State of Louisiana P. O. Box 94005 Baton Rouge, LA 70804-9005 Ms. H. Anne Plettinger 3456 Villa Rose Drive Baton Rouge, LA 70806 President of West Feliciana Police Jury P. O. Box 1921 St. Francisville, LA 70775 Mr. Brian Almon Public Utility Commission William B. Travis Building P. O. Box 13326 Austin, TX 78701-3326 Mr. Jim Calloway Public Utility Commission of Texas 1701 N. Congress Avenue Austin, TX 78711-3326 Louisiana Department of Environmental Quality Radiological Emergency Planning and Response Division P. O. Box 4312 Baton Rouge, LA 70821-4312 Joseph A. Aluise Associate General Counsel - Nuclear Entergy Services, Inc. 639 Loyola Avenue New Orleans, LA 70113 Chief, Technological Hazards Branch FEMA Region VI 800 N. Loop 288 Denton, TX 76209-3606 Entergy Operations, Inc. - 4 -

Electronic distribution by RIV: Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Chuck.Casto@nrc.gov) DRP Acting Director (Anton.Vegel@nrc.gov)

DRP Acting Deputy Director (Troy.Pruett@nrc.gov) DRS Director (Roy.Caniano@nrc.gov) DRS Deputy Director (Jeff.Clark@nrc.gov) Senior Resident Inspector (Grant.Larkin@nrc.gov) Resident Inspector (Charles.Norton@nrc.gov)

Branch Chief, DRP/C (Vincent.Gaddy@nrc.gov) RBS Administrative Assistant (Lisa.Day@nrc.gov) Senior Project Engineer, DRP/C (Bob.Hagar@nrc.gov) Project Engineer, DRP/C (Rayomand.Kumana@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Alan.Wang@nrc.gov)

Branch Chief, DRS/TSB (Michael.Hay@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) OEMail Resource ROPreports DRS/TSB STA (Dale.Powers@nrc.gov) OEDO RIV Coordinator (Margie.Kotzalas@nrc.gov) W. A. Maier, RSLO (Bill.Maier@nrc.gov) E. P. Schrader, NSIR/DPR/EP (Eric.Schrader@nrc.gov)

File located: R:\_REACTOR\_RB\2010\RB 2010003-RP-GFL.docx ML 102170526 SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials VGGPublicly Avail Yes No Sensitive Yes No Sens. Type Initials VGGRI:DRP/ SRI:DRP/ C:DRS/EB1 C:DRS/EB2 CHNorton GFLarkin TRFarnholtz NFO'Keefe

/VGG for/ /VGG for/ /RA//STG for/ 8/4/10 8/4/10 7/20/10 7/20/10 C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRS/TSB C:DRP/C MSHaire MPShannon GEWerner MCHay VGGaddy

/BTL for/ /DAP for/ /RA/ /DAP for/ 7/22/10 7/20/10 7/23/10 7/20/10 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 05000458 License: NPF-47 Report: 05000458/2010003 Licensee: Entergy Operations, Inc. Facility: River Bend Station Location: 5485 U.S. Highway 61N St. Francisville, LA Dates: April 1 through June 30, 2010 Inspectors: G. Larkin, Senior Resident Inspector C. Norton, Resident Inspector A. Fairbanks, Reactor Inspector, Engineering Branch 1 P. Elkmann, Senior Emergency Preparedness Inspector L. Carson II, Senior Health Physicist, Plant Support Branch 2 N. Greene, Health Physicist, Plant Support Branch 2 Approved By: Vincent G. Gaddy, Chief, Project Branch C Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000458/2010003; 04/01/2010 - 06/30/2010; River Bend Station; Integrated Inspection Report; Maintenance Risk Assessments and Emergent Work Control; Operability Evaluations; Occupational Dose Assessment

The report covered a 3-month period of inspection by resident inspectors and announced baseline inspections by regional based inspectors. Three Green noncited violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensee's failure to perform an adequate risk assessment while the high pressure core spray room unit cooler was unavailable. Specifically, the licensee assumed that risk would remain green and high pressure core spray would continue to inject into the reactor vessel for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after room cooling was made unavailable, when, in fact, risk became yellow because high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pump's minimum flow logic. As immediate corrective action, the licensee issued a standing order that administratively considered high pressure core spray unavailable when its room cooler is removed from service. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02937.

This issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection 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 and because the licensee used incorrect risk assumptions that changed the outcome of their risk assessment. There is no crosscutting aspect associated with this violation because the assumptions that lead to the performance deficiency are not indicative of current licensee performance (Section 1R13).

Green.

A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for inadequate procedural guidance when surveillance testing the Division III diesel generator. This resulted in a loss of offsite power to the Division III 4160 volt alternating current (Vac) bus while starting a nonsafety-related load with the Division III emergency diesel generator at full power paralleled to the grid. To prevent isolating offsite power to any safety bus, the licensee issued procedure changes to prevent starting loads on the nonsafety-related buses connected to the divisional safety buses while the emergency diesel generators are in test. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-00910.

The finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Table 4a, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its technical specification allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk significant, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. There is no crosscutting aspect associated with this violation because this is a historical condition not previously identified by the licensee (Section 1R15).

Cornerstone: Occupational Radiation Safety

Green.

NRC inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 for failure to follow radiation work permit instructions. Specifically, a team technician made an unauthorized entry into a posted high radiation area on a radiation work permit that did not grant access to that area. The licensee conducted a review of this event and issued a site-wide memorandum on procedural and management expectations associated with high radiation areas. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2009-03953.

The failure to follow the instructions on a radiation work permit is a performance deficiency. The performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone. It is associated with the exposure control attribute in that a worker not following radiation work permit instructions does not ensure adequate protection of the worker's health and safety from additional/unintended personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Furthermore, the finding had an associated human performance crosscutting aspect in the work practices component because the worker did not use human error prevention techniques, such as self-checking H.4(a)(Section 2RS04).

B. Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors. Corrective actions taken by the licensee have been entered into the licensee's corrective action program. This violation and corrective actions are listed in Section 40A7 of this report.

REPORT DETAILS

Summary of Plant Status

River Bend Station began the inspection period at 100 percent thermal power. On April 10, 2010, the plant reduced reactor power to 95 percent to exercise partially withdrawn control rods and perform turbine bypass valve testing. The plant returned to full power on April 10, 2010. On May 7, 2010, the plant reduced reactor power to 95 percent to exercise partially withdrawn control rods and perform turbine bypass valve testing. The plant returned to full power on May 7, 2010. On June 1, 2010, the plant reduced reactor power to 70 percent to remove reactor feedwater pump 1A from service for shaft seal replacement. The plant returned to 100 percent reactor power on June 7, 2010, and remained at 100 percent reactor power for the remainder of the inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Summer Readiness for Offsite and Alternate-ac Power

a. Inspection Scope

The inspectors performed a review of preparations for summer weather for selected systems, including conditions that could lead to loss-of-offsite power and conditions that could result from high temperatures. The inspectors reviewed the procedures affecting these areas and the communications protocols between the transmission system operator and the plant to verify that the appropriate information was being exchanged when issues arose that could affect the offsite power system. Examples of aspects considered in the inspectors' review included:

• The coordination between the transmission system operator and the plant's operations personnel during off-normal or emergency events

• The explanations for the events

• The estimates of when the offsite power system would be returned to a normal state

• The notifications from the transmission system operator to the plant when the offsite power system was returned to normal

During the inspection, the inspectors focused on plant-specific design features and the procedures used by plant personnel to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Safety Analysis Report and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors' reviews focused specifically on the following plant systems:

• Fancy Point 500 kV and 230kV substation

• River Bend transformer yard These activities constitute completion of one readiness for summer weather affect on offsite and alternate-ac power sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings were identified.

.2 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors performed a review of the adverse weather procedures for seasonal extreme low temperatures. The inspectors verified that weather-related equipment deficiencies identified during the previous year were corrected prior to the onset of seasonal extremes, and evaluated the implementation of the adverse weather preparation procedures and compensatory measures for the affected conditions before the onset of, and during, the adverse weather conditions.

During the inspection, the inspectors focused on plant-specific design features and the procedures used by plant personnel to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Safety Analysis Report and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective action program items to verify that plant personnel were identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors' reviews focused specifically on the following plant systems:

• Control building chilled water system

• Standby service water system These activities constitute completion of one readiness for seasonal adverse weather sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignments

Partial Walkdown

a. Inspection Scope

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

• Reactor core isolation cooling

• Division II emergency diesel generator

• Main steam positive leakage control system

The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report, technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also inspected accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

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

• April 4, 2010, diesel generator building, 98-foot elevation; control building, 98-foot elevation

• April 18, 2010, reactor core isolation cooling pump room

• April 28, 2010, control building, 116-foot elevation and 136-foot elevation

• May 13, 2010, auxiliary building, crescent area; E, F, and G tunnels; low pressure core spray pump room; auxiliary building 114-foot elevation and 141-foot elevation

• May 31, 2010, auxiliary building, 141-foot elevation; low pressure core spray pump room; Division I emergency diesel room

• June 8, 2010, fire pump house, 95-foot elevation

• June 21, 2010, control building, 70-foot elevation and 98-foot elevation The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensee's fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plant's ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensee's corrective action program. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of seven quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings were identified.

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On April 20, 2010, the inspectors observed a fire brigade activation for a fire in the Division II switchgear room. The observation evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were (1) proper wearing of turnout gear and self-contained breathing apparatus; (2) proper use and layout of fire hoses; (3) employment of appropriate fire fighting techniques; (4) sufficient firefighting equipment brought to the scene; (5) effectiveness of fire brigade leader communications, command, and control; (6) search for victims and propagation of the fire into other plant areas; (7) smoke removal operations; (8) utilization of preplanned strategies; (9) adherence to the preplanned drill scenario; and (10) drill objectives.

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

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

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

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

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On June 22, 2010, and June 29, 2010, the inspectors observed a crew of licensed operators in the plant's simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crew's clarity and formality of communications

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

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

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

• Control board manipulations

• Oversight and direction from supervisors

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

These activities constitute completion of two quarterly licensed-operator requalification program samples as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

R12 Maintenance Effectiveness

a. Inspection Scope

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

• Control building chilled water system The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

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

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

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

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly maintenance effectiveness sample as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

• Transmission and distribution communication work in the Fancy Point 230kV switchyard, April 8, 2010

• Monitoring of transmission and distribution work on line 354, April 12, 2010

• Actions to mitigate steam tunnel high temperature, May 14, 2010

• Division II standby gas treatment system heater failure, May 14, 2010

• Main steam line drains emergent maintenance, May 21, 2010

• Division I main steam positive leakage control system emergent maintenance, May 28, 2010

• Division I standby service water fan cell emergent maintenance, June 3, 2010

• High pressure core spray room unit cooler emergent maintenance, June 28, 2010 The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of eight maintenance risk assessments and emergent work control inspection samples as defined in Inspection Procedure 71111.13-05.

b. Findings

Introduction.

The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) involving the licensee's failure to perform an adequate risk assessment while the high pressure core spray room unit cooler was unavailable. Specifically, the licensee assumed that risk would remain green (low) and high pressure core spray would continue to inject into the reactor vessel for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, after room cooling was made unavailable, when in fact risk became yellow (elevated) because high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pump's minimum flow logic. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02937.

Description.

On June 25, 2010, the licensee removed the high pressure core spray pump room unit cooler from service for emergent maintenance. The licensee assessed the plant risk as green using their Equipment-Out-Of-Service risk model. The inspectors challenged the model's assumption that high pressure core spray was available for accident mitigation without heat removal support from its room cooler. Without technical basis, the probabilistic safety analysis assumed that high pressure core spray would continue to inject into the reactor vessel for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after loss of room cooling when, in fact, high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pump's minimum flow logic that could cause the pump to become dead headed, overheat, and mechanically fail. During an accident scenario, the probabilistic safety analysis required high pressure core spray to provide core cooling for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following plant shutdown at which time the control rod drive pumps would have the capacity to provide adequate core cooling. With the high pressure core spray pump room cooler unavailable, the probabilistic safety analysis fault tree inappropriately modeled high pressure core spray as available for all accident scenarios other than station blackout. This modeling error was originally made in the early 1990s. The licensee determined with high pressure core spray unavailable to perform its mission that the plant safety index was 8.4 (yellow) instead of 9.5 (green). As immediate corrective action, the licensee issued a standing upgrade that administratively considered high pressure core spray unavailable when its room cooler is removed from service.

Analysis.

The inspectors determined that the licensee's failure to perform an adequate risk assessment was a performance deficiency. Specifically, the licensee's risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. This issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and because the licensee used incorrect risk assumptions that changed the outcome of their risk assessment. Using Inspection 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. There is no crosscutting aspect associated with this violation because the assumptions that lead to the performance deficiency are not indicative of current licensee performance.

Enforcement.

Title 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, on June 25, 2010, the licensee failed to perform an adequate risk assessment before performing maintenance on the high pressure core spray room unit cooler. Since an adequate risk assessment was not performed, the licensee failed to manage the actual risk associated with the activity. Because the finding was of very low safety significance and it has been entered into the licensee's corrective action program as Condition Report CR-RBS-2010-02937, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000458/2010003-01, "Inadequate Risk Assessment for High Pressure Core Spray Room Unit Cooler Maintenance."

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• CR-RBS-2010-00910, unplanned trip of Division III 4160 Vac tie breaker, reviewed on May 4, 2010

• CR-RBS-2010-01785, reactor core isolation cooling flow indication, reviewed on May 14, 2010

• CR-RBS-2010-01717, reactor core isolation cooling low oil level, reviewed on May 14, 2010

• CR-RBS-2010-02468, reported main steam positive leakage control system pipe flaw on ASME Code class 1 piping, reviewed on May 28, 2010

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and Updated Safety Analysis Report to the licensee personnel's evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four operability evaluations inspection samples as defined in Inspection Procedure 71111.15-04

b. Findings

Introduction.

A Green self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for inadequate procedural guidance when surveillance testing the Division III diesel generator. This resulted in a loss of offsite power to the Division III 4160 volt alternating current (Vac) bus while starting a nonsafety-related load with the Division III emergency diesel generator at full power paralleled to the grid. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-00910.

Description.

On February 24, 2010, while running the Division III emergency diesel generator at full load and parallel to the grid per surveillance test procedure STP-309-0203, "DIV 3 Diesel Generator Operability Test," the 4160 Vac safety-related to nonsafety-related tie breaker (E22-S004-ACB04) tripped on directional over current. The breaker is located electrically between the Division III safety bus and both technical specification required offsite power circuits. The breaker trip resulted in the loss of offsite power to the Division III safety bus and made high pressure core spray and standby service pump 2C inoperable. The directional over current protection causes the breaker to trip and separate the Division III safety bus from offsite power when the diesel is aligned in parallel with the offsite power and excess current flow from the safety-related bus longer than a prescribed time delay. Review of computer data points indicated at the instant before the breaker tripped, operators started the 1,250 horsepower turbine building ventilation chiller motor 1B, per the approved work week schedule. There was insufficient procedural guidance to limit starting large electrical loads while testing emergency diesel generators. This was a performance deficiency. A contributing cause was the failure to consider the load impact of starting HVN-CHL1B during surveillance test scheduling. To prevent isolating offsite power to any safety bus, the licensee issued procedure changes to prevent starting large loads on the nonsafety-related busses connected to the divisional safety busses while the emergency diesel generators are in test.

Analysis.

The failure to provide adequate procedural guidance to prevent the loss of offsite power to the Division III 4160 Vac safety bus during periodic testing was a performance deficiency. The finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Table 4a, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its technical specification allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk significant, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. There is no crosscutting aspect associated with this violation because this is a historical condition not previously identified by the licensee and is not reflective of present licensee performance.

Enforcement.

Technical Specification 5.4.1.a. requires that written procedures be established, implemented, and maintained covering the activities in NRC Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2. Regulatory Guide 1.33, Appendix A, Section 4.w, requires procedures be prepared, as appropriate, for operation of safety-related systems. Contrary to the above, STP-309-0203, "DIV 3 Diesel Generator Operability Test," did not provide guidance to prevent starting loads during emergency diesel generator surveillance testing that could result in the loss of offsite power. On February 24, 2010, the licensee started the nonsafety-related turbine building ventilation chiller which tripped the safety to nonsafety-related tie breaker and caused a loss of all offsite power to Division III safety-related busses. This resulted in an unplanned inoperability of high pressure core spray and standby service water pump 2C for greater than 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />. Because this violation is of very low safety significance and it has been entered into the licensee's corrective action program as Condition Report CR-RBS-2010-00910, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000458/2010003-02, "Inadequate Procedure Results in Loss of Offsite Power to a 4160 Vac Safety Bus."

1R18 Plant Modifications

.1 Temporary Modifications

a. Inspection Scope

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

• Engineering Change EC-21957, "Temporary Change to Main Steam Line Tunnel Ambient Temperature High Trip Set Point," Revision 0

• Engineering Change EC-22658, "E51-MOVF064, RCIC Control Isolation Valve, Evaluation to Increase the Allowable Stroke Time," Revision 0 The inspectors reviewed the temporary modifications and the associated safety evaluation screening against the system design bases documentation, including the Updated Safety Analysis Report and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors verified that the temporary modifications were identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

These activities constitute completion of two samples for temporary plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings were identified.

.2 Permanent Modifications

The inspectors reviewed key affected parameters associated with energy needs, materials, replacement components, timing, heat removal, control signals, equipment protection from hazards, operations, flow paths, pressure boundary, ventilation boundary, structural, process medium properties, licensing basis, and failure modes for the permanent modifications listed below.

• Engineering Change EC-2570, "Provide an Alternate Power Source for E51-MOVF063 During a Main Control Room Fire," Revision 0

• Engineering Change EC-11873, "NobleChem Application Procedure for River Bend Station, Revision 0 The inspectors verified that modification preparation, staging, and implementation did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; postmodification testing will maintain the plant in a safe configuration during testing by verifying that unintended system interactions will not occur; systems, structures and components' performance characteristics still meet the design basis; the modification design assumptions were appropriate; the modification test acceptance criteria will be met; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two samples for permanent plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings were identified.

R19 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 00228482, "MSS-HYVCV2 Fast Acting Solenoid Did Not Actuate," reviewed on May 17, 2010

• WO 00238997, "Electrically Backseat E51-MOVF064 In Accordance With EC-22647 Requirements," reviewed on June 2, 2010

• WO 00239242, "Postmaintenance Test of Division I Standby Service Water Fan L Time Start Delay Relay Replacement," reviewed on June 2, 2010 The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following (as applicable):

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

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate

The inspectors evaluated the activities against the technical specifications, the Updated Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

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

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Procedures

• Test data

• Testing frequency and method demonstrated technical specification operability

• Restoration of plant systems

• Reference setting data

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

• OSP-0102, "Turbine Valve Testing," performed on April 7, 2010

• STP-256-0203, "Division I Cooling Tower Fans Operability Test," performed on June 2, 2010

• STP-000-0001, "Daily Operating Logs (for unidentified RCS leakage)," performed on June 8, 2010 Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three surveillance testing inspection samples as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed an in-office review of River Bend Station Procedure EIP-2-001, "Classification of Emergencies," Revision 18. This revision replaced the definition of 'hostile action' with the definition of Nuclear Energy Institute Report 99-01, "Emergency Action Level Methodology," Revision 5, and defined 'security condition' and 'imminent' as used in evaluating emergency action levels. The revision replaced security-related emergency action levels HG1, HS1, HA1, and HU1, and their associated technical bases, with the corresponding emergency action levels from NEI 99-01, "Emergency Action Level Methodology," Revision 5, and made other minor editorial corrections.

This revision was compared to its previous revision, to the criteria of NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, to Nuclear Energy Institute 99-01, "Emergency Action Level Methodology," Revisions 4 and 5, 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 licensee-generated changes; therefore, this revision is subject to future inspection.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.04-05.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

Training Observations

a. Inspection Scope

The inspectors observed a simulator training evolution for licensed operators on June 22, 2010, which required emergency plan implementation by a licensee operations crew. This evolution was planned to be evaluated and included in performance indicator data regarding drill and exercise performance. The inspectors observed event classification and notification activities performed by the crew. The inspectors also attended the post-evolution critique for the scenario. The focus of the inspectors' activities was to note any weaknesses and deficiencies in the crew's performance and ensure that the licensee evaluators noted the same issues and entered them into the corrective action program. As part of the inspection, the inspectors reviewed the scenario package and other documents listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.06-05.

b. Findings

No findings were identified.

RADIATION SAFETY

[RS] Cornerstone: Occupational and Public Radiation Safety

2RS0 4 Occupational Dose Assessment

a. Inspection Scope

This area was inspected to: (1) determine the accuracy and operability of personal monitoring equipment; (2) determine the accuracy and effectiveness of the licensee's methods for determining total effective dose equivalent; and (3) ensure occupational dose is appropriately monitored. The inspectors interviewed licensee personnel and reviewed the following:

• Self-assessments, audits, corrective action documents, procedures, and reports related to the dosimetry program since the last inspection

• External dosimetry accreditation, storage, issue, use, and processing of active and passive dosimeters

• The technical competency and adequacy of the licensee's internal dosimetry program

• Adequacy of the dosimetry program for special dosimetry situations such as declared pregnant workers, multiple dosimetry placement, and neutron dose assessment Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of the one required sample as defined in Inspection Procedure 71124.04-05.

b. Findings

Introduction.

NRC inspectors reviewed a Green self-revealing, noncited violation of Technical Specification 5.4.1 for failure to follow the radiation work permit instructions while working in the feedwater pump area of the turbine building.

Description.

On September 3, 2009, the Fix-It-Now team technician was dispatched to the turbine building to investigate the cause of a steam seal evaporator problem. The technician logged into the radiologically controlled area on Task 1 of Radiation Work Permit 20091012, Revision 01. This radiation work permit granted the team technician access to all areas of the turbine building that were less than 100 mrem per hour at 30 centimeters. The team technician was not allowed to access high radiation areas or locked high radiation areas. However, the technician entered the 67 foot elevation of the turbine building to observe a controller and obtain a reading. This specific area was posted as a high radiation area. As a result of entering the posted high radiation area, the technician received an electronic dosimeter dose rate alarm. However, the technician failed to hear the alarm due to background noise. This entry resulted in a maximum dose rate of 131 mrem per hour. The dose rate alarm setpoint was 80 mrem per hour. The technician's total dose received during the tour was 6 mrem. The survey record of the feedwater pump area within the turbine building revealed a maximum dose rate of 160 mrem per hour. The licensee's response to this violation was to conduct a human performance error review and initiate a condition report. Additionally, a memorandum was issued to the site and the Entergy fleet re-enforcing the procedural requirements and expectations for entering high radiation areas.

Analysis.

The failure to follow a radiation work permit instruction is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety cornerstone. It is associated with the exposure control attribute in that a worker not following radiation work permit instructions does not ensure adequate protection of the worker's health and safety from additional/unintended personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Furthermore, the finding had an associated human performance cross-cutting aspect in the work practices component because the worker did not use human error prevention techniques, such as self-checking H.4(a).

Enforcement.

Technical Specification 5.4.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Section 7(e)(1), "Radiation Protection Procedures," of Appendix A to Regulatory Guide 1.33 lists procedures for access control to radiation areas, including a radiation work permit system. Procedure EN-RP-105, "Radiation Worker Permits," Revision 8, Section 4.0[5], states that the radiation worker is responsible for reviewing the radiation work permit and complying with the requirements. Radiation Work Permit 20091012, Task 1, did not permit entry into high radiation areas. Contrary to the above, on September 3, 2009, the team technician did not comply with all written instructions provided relative to his radiation work permit, in that he inappropriately entered a high radiation area.

Considering this failure to follow procedural guidance when entering the radiologically controlled area was of very low safety significance and it has been entered into the licensee's corrective action program in Condition Report CR-RBS-2009-03953, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000458/2010003-03 "Failure to Follow Radiation Work Permit Instructions."

2RS0 5 Radiation Monitoring Instrumentation

a. Inspection Scope

To verify that the licensee is assuring the accuracy and operability of radiation monitoring instruments that are used to: (1) monitor areas, materials, and workers to ensure a radiologically safe work environment; and (2) detect and quantify radioactive process streams and effluent releases. The instrumentation subject to this review includes equipment used to monitor radiological conditions incident to normal plant operations, including anticipated operational occurrences, and from postulated accidents. The team used the requirements in 10 CFR Part 20 and the licensee's procedures required by technical specifications as criteria for determining compliance. The inspectors interviewed licensee personnel and reviewed the following:

• Selected plant configurations and alignments of process, postaccident, and effluent monitors with descriptions in the Updated Final Safety Analysis Report and the offsite dose calculation manual

• Select instrumentation, including effluent monitoring instrument, portable survey instruments, area radiation monitors, continuous air monitors, personnel contamination monitors, portal monitors, and small article monitors to examine their configurations and source checks

• Calibration and testing of process and effluent monitors, laboratory instrumentation, whole body counter, postaccident monitoring instrumentation, portal monitors, personnel contamination monitors, small article monitors, portable survey instruments, area radiation monitors electronic dosimetry, and air samplers/continuous air monitors

• Reports, self-assessments, and corrective action reports performed since the last inspection related to radiation monitoring instrumentation and follow-up activities, such associated with problem identification, characterization, and tracking Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of the one required sample as defined in Inspection Procedure 71124.05-05.

b. Findings

No findings were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the performance indicator data submitted by the licensee for the second quarter 2010 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, "Performance Indicator Program."

This review was performed as part of the inspectors' normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings were identified.

.2 Safety System Functional Failures (MS05)

a. Inspection Scope

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

These activities constitute completion of one safety system functional failures sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

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

a. Inspection Scope

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

These activities constitute completion of one mitigating systems performance index emergency ac power system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.4 Mitigating Systems Performance Index - High Pressure Injection Systems (MS07)

a. Inspection Scope

The inspectors sampled licensee submittals for the mitigating systems performance index - high pressure injection systems performance indicator for the period from the second quarter 2009 through the first quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6. The inspectors reviewed the licensee's operator narrative logs, issue reports, mitigating systems performance index derivation reports, event reports, and NRC integrated inspection reports for the period of April 2009 through March 2010 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one mitigating systems performance index high pressure injection system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

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

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

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

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

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's corrective action program. The inspectors accomplished this through review of the station's daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensee's corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening discussed in Section 4OA2.2, licensee trending efforts, and licensee human performance results. The inspectors nominally considered the 6-month period of October 1, 2009, through March 31, 2010; although, some examples expanded beyond those dates where the scope of the trend warranted.

The inspectors also included issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments. The inspectors compared and contrasted their results with the results contained in the licensee's corrective action program trending reports. Corrective actions associated with a sample of the issues identified in the licensee's trending reports were reviewed for adequacy.

These activities constitute completion of one single semi-annual trend inspection sample as defined in Inspection Procedure 71152-05.

b. Findings

No findings were identified.

.4 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensee's corrective action program, the inspectors recognized a corrective action item requiring an equipment failure evaluation concerning:

. The large number of steam leaks (15) following start-up from the refuel outage in mid-October 2009. Two through wall pipe flaws and 13 mechanical joints leaks were reviewed to determine if additional actions were required to maintain plant safety. Six steam leaks were associated with motor operated valve leak off lines. The apparent cause for these leaks was failure to include adequate work instructions to install thread sealant on the leak off line threads prior to reinstallation. Two of the three drain valve leaks stopped after the valves were taken fully shut. In addition the pipe wall flaws were evaluated in the flow accelerated corrosion program for expanded inspections in the areas where they were detected.

2. The inspectors focused on the emergency diesel generators' vibration issues, associated with the combustion air intake system where fatigue cracking have contributed to structural failures since 1989, to determine if the licensee's corrective actions were properly focused on each structural failure and addressed the global cause of the failures.

These activities constitute completion of two in-depth problem identification and resolution sample defined in Inspection Procedure 71152-05.

4OA3 Event Followup

.1 (Closed) Licensee Event Report 05000458/2009003-00: Reactor Pressure Trip Unit

Inoperable Greater than the Allowed Outage Time

From April 11, 2008, until June 7, 2008, one channel of reactor pressure instrumentation for the anticipated transient without scram recirculation pump trip was outside the technical specification allowed set point range. This occurred because the test instrument used to calibrate the reactor pressure instrumentation on April 11 was out of calibration. This went unrecognized until the next scheduled calibration on June 7. The licensee determined that job performance standards were not adequately defined for the job. Corrective actions included better defining performance standards for the job. This finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in the loss of operability functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train from greater than its technical specification allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk-significant per 10 CFR50.65 for greater that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and was not potentially risk significant due to seismic, flooding, or severe weather initiating event. This licensee identified violation involved a violation of Technical Specification 3.3.4.2.

.2 (Closed) Licensee Event Report 05000458/2010-001-00:

Control Building Chiller Inoperable Greater Than Allowable Outage Time This licensee event report discusses that during accident conditions the control building chillers were not able to remove the design basis heat load while operating with low standby cooling water temperatures. The inspectors identified this issue as a Green noncited violation of Technical Specification 3.7.3 for exceeding the control room air conditioning system thirty day allowed outage time for one inoperable subsystem, the seven day allowed outage time for two inoperable subsystems and failing to enter Modes 3 and 4, as specified. See NRC Inspection Report 05000458/2010002 for additional details. This licensee event report is closed.

4OA6 Meetings

Exit Meeting Summary

On May 10, 2010, the inspector conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensee's emergency plan implementing procedure to Mr. T. Burnett, Manager, Emergency Preparedness, and other members of the licensee's staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On May 27, 2010, the inspectors presented the radiation safety inspection results to Mr. M. Perito, Site Vice President, Operations, and other members of the licensee staff. The licensee staff acknowledged the issues presented. The inspector asked the licensee staff whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On July 8, 2010, the inspectors presented the integrated inspection results to Mr. B. Cox, Acting Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited violation.

Technical Specification 3.3.4.2 allows 14 days inoperability of one channel of the anticipated transient without scram recirculation pump trip instrumentation. Contrary to this requirement one channel of instrumentation was outside the technical specification allowed set point range from April 11, 2008, to June 7, 2008. This was identified by the licensee and entered into the licensee's corrective action program as Condition Report CR-RBS-2008-03719. This finding was determined to have very low safety significance.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Burnett, Manager, Emergency Preparedness

G. Bush, Manager, Maintenance

M. Caffervel, Auditor, Quality Assurance

M. Chase, Manager, Training

J. Clark, Assistant Operations Manager - Shift

B. Cox, Manager, Operations

M. Feltner, Manager, Outage

C. Forpahl, Manager, Engineering Programs & Components

W. Fountain, Senior Licensing Specialist

H. Goodman, Director, Engineering

D. Heath, Supervisor, Radiation Protection

R. Heath, Manager, Chemistry

B. Houston, Manager, Radiation Protection

K. Huffstatler, Senior Licensing Specialist

A. James, Manager, Security

M. Jureg, Auditor, Quality Assurance

L. Kitchen, Manager, Planning, Scheduling, and Outages

R. Kowalewski, Manager, Corrective Actions & Assessments

G. Krause, Assistant Operations Manager - Support

D. Lorfing, Manager, Licensing

W. Mashburn, Manager, Design Engineering

R. McAdams, Manager, System Engineering

E. Olson, General Manager, Plant Operations

S. King-Patterson, Radiation Protection Technician

M. Perito, Site Vice President

R. Persons, Superintendent, Training

J. Roberts, Director, Nuclear Safety Assurance

T. Shenk, Assistant Operations Manager - Training

J. Vollmer, Senior Health Physicist/Chemistry Specialist

J. Vukovics, Supervisor, Reactor Engineering

J. Wilson, Supervisor, Mechanical Systems

L. Woods, Manager, Quality Assurance

NRC Personnel

G. Larkin, Senior Resident Inspector

C. Norton, Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000458/2010003-01 NCV Inadequate Risk Assessment for High Pressure Core Spray Pump Room Unit Cooler Maintenance

(Section 1R13)

05000458/2010003-02 NCV Inadequate Procedure Results in Loss of Offsite Power to a 4160 Vac Safety Bus (Section 1R15)

05000458/2010003-03 NCV Failure to Follow Radiation Work Permit Instructions (Section 2RS04)

Closed

05000458/LER-2010-001-00 LER Control Building Chiller Inoperable Greater Than Allowable Outage Time (Section 4OA3)

05000458/LER-2008-003-02 LER Reactor Pressure Trip Unit Inoperable Greater Than the Allowable Outage Time (Section 40A3)

Section 1R01: Adverse Weather Protection

CALCULATION

NUMBER TITLE REVISION G13.18.2.1*018 Control Building Air Handling Units Chilled Water Flow Requirements and Associated Required Chiller Capacity 1

CONDITION REPORTS

CR-RBS-2008-01415

CR-RBS-2010-00692

CR-RBS-2009-04204

CR-RBS-2009-05795

CR-RBS-2010-00239

Attachment

MISCELLANEOUS

NUMBER TITLE REVISION/DATE

Licensee Event Report 94-005-01 Loss of Both Trains of Control Room Emergency Recirculation Due to Low Emergency Closed Cooling Temperature October 28, 1994 Regulatory Guide 1.160 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants

PROCEDURES

NUMBER TITLE REVISION

ADM-0096 Risk Management Program and Implementation Risk Assessment

304

AOP-0064 Degraded Grid 4

EN-DC-204 Maintenance Rule Scope and Basis 2

EN-DC-205 Maintenance Rule Monitoring 2

EN-LI-114 Performance Indicator Process 3

EN-OP-104 Operability Determination Process 4

ENS-DC-199 Off-Site Power Supply Design Requirements 2

ENS-DC-201 Transmission Grid Monitoring 2

OSP-0028 Log Report - Normal Switchgear, Control, and Diesel Generator Buildings

055

OSP-0063 Grid Monitor 3 R-STM-0402 HVAC - Control Building and Diesel Generator Building 4

SDC-402/410 Control Building HVAC System Control Building Chilled Water System Ventilation Chilled Water System Design Criteria System Numbers 402, & 410

Attachment

Section 1R04: Equipment Alignment

PROCEDURES

NUMBER TITLE REVISION

SOP-0035 Reactor Core Isolation Cooling System 35

SOP-0053 Standby Diesel Generator and Auxiliaries 310

STP-208-6302 DIV II MSIV Leakage Control Quarterly Valve Operability Test 6

Section 1R05: Fire Protection

PROCEDURE

NUMBER TITLE REVISION

AOP-0052 Fire Outside the Main Control Room in Areas Containing Safety Related Equipment

SCENARIO

NUMBER TITLE REVISION

DRL-FP-0106 Fire in the Division III Switchgear Room 0

Section 1R07: Heat Sink Performance

CALCULATION

NUMBER TITLE REVISION G13.18.2.1*018 Control Building Air Handling Units Chilled Water Flow Requirements and Associated Required Chiller Capacity 1

CONDITION REPORTS

CR-RBS-2008-01415

CR-RBS-2010-00692

CR-RBS-2009-04204

CR-RBS-2009-05795

CR-RBS-2010-00239

Attachment

MISCELLANEOUS

NUMBER TITLE REVISION/DATE

Licensee Event Report 94-005-01 Loss of Both Trains of Control Room Emergency Recirculation Due to Low Emergency Closed Cooling Temperature October 28, 1994 Regulatory Guide 1.160 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants

PROCEDURES

NUMBER TITLE REVISION

ADM-0096 Risk Management Program and Implementation Risk Assessment

304

EN-DC-204 Maintenance Rule Scope and Basis 2

EN-DC-205 Maintenance Rule Monitoring 2

EN-LI-114 Performance Indicator Process 3

EN-OP-104 Operability Determination Process 4

OSP-0028 Log Report - Normal Switchgear, Control, and Diesel Generator Buildings

055 R-STM-0402 HVAC - Control Building and Diesel Generator Building 4

SDC-402/410 Control Building HVAC System Control Building Chilled Water System Ventilation Chilled Water System Design Criteria System Numbers 402 & 410

Section 1R11: Licensed Operator Requalification Program

SCENARIOS

NUMBER TITLE REVISION

RSMS-OPS-635 APRM F, Feed Pump Min Flow Valve, RWCU Leak - ATWS 0

Attachment NUMBER TITLE REVISION

RSMS-OPS-636 RPV Level Instrument Failure, Single Rod Scram, Loss of Turbine Oil, ATWS

Section 1R12: Maintenance Effectiveness

CONDITION REPORTS

CR-RBS-2008-01246

CR-RBS-2008-05391

CR-RBS-2009-01864

CR-RBS-2009-02023

CR-RBS-2009-02882

CR-RBS-2009-04204

CR-RBS-2009-06063

CR-RBS-2010-00239

CR-RBS-2010-00822

CR-RBS-2008-01286

CR-RBS-2009-01140

CR-RBS-2009-01873

CR-RBS-2009-02042

CR-RBS-2009-02963

CR-RBS-2009-04598

CR-RBS-2009-06147

CR-RBS-2010-00547

CR-RBS-2010-00866

CR-RBS-2008-02206

CR-RBS-2009-01271

CR-RBS-2009-01886

CR-RBS-2009-02147

CR-RBS-2009-03001

CR-RBS-2009-04722

CR-RBS-2009-06204

CR-RBS-2010-00738

CR-RBS-2010-01667

CR-RBS-2008-05229

CR-RBS-2009-01539

CR-RBS-2009-01967

CR-RBS-2009-02691

CR-RBS-2009-03911

CR-RBS-2009-05443

CR-RBS-2010-00237

CR-RBS-2010-00799 MAINTENANCE RULE DOCUMENT

Maintenance Rule Program 2007-08 (a)(3) Periodic Assessment Engineering Report #RBS-SE-09-00001, Revision 000

MISCELLANEOUS

Calculation "RBS Revision 4 PSA Summary Report," Document Number

PRA-RB-01-002

PROCEDURES

NUMBER TITLE REVISION

EN-DC-205

EN-DC-206

EN-DC-207 Maintenance Rule Monitoring Maintenance Rule (a)(1) Process Maintenance Rule Periodic Assessment

1 2

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

CALCULATIONS

NUMBER TITLE REVISION G13.18.2.7*22 Steam Tunnel leak Detection Temperature Limit Sensitivity to Initial Temperature and Leak Rate

Attachment NUMBER TITLE REVISION G13.18.6.1-E31*009 Instrument Loop Uncertainty/Setpoint Determination for Main Steam Line Tunnel Ambient Temperature - High 0

CONDITION REPORTS

CR-RBS-2004-01752

CR-RBS-2010-02337

CR-RBS-2010-02937

CR-RBS-2009-06036

CR-RBS-2010-02468

CR-RBS-2010-01981

CR-RBS-2010-02503

CR-RBS-2010-02045

CR-RBS-2010-02530

DRAWING

NUMBER TITLE REVISION

TLD-LMS-075 Test Loop Diagram Main Steam Line Tunnel Ambient Temperature E31-T/CN031A

MISCELLANEOUS

TITLE REVISION ODMI Operation with Main Steam Tunnel High Ambient Temperature

PROCEDURES

NUMBER TITLE REVISION

ADM-0096 Risk Management Program Implementation Risk Assessment 305/306

ARP-601-19 P601-19 Alarm Response 026

EN-MA-125 Troubleshooting Control of Maintenance Activities 6

EOP-0003 Emergency Operating Procedure - Secondary Containment and Radioactive Release Control

014

OSP-0048 Switchyard, Transformer Yard, and Sensitive Equipment Controls 007

Attachment TECHNICAL SPECIFICATIONS

NUMBER TITLE AMENDMENT

Technical Specification 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation

165 Technical Specification 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 81

UPDATED SAFETY ANALYSIS REPORT DOCUMENT

NUMBER TITLE REVISION USAR 10.3 Main Steam Supply System 14

WORK ORDER

NUMBER TITLE

WO 239242 Replace Service Water Fan 1L Time Delay Relay 62L

Section 1R15: Operability Evaluations

CALCULATIONS

NUMBER TITLE REVISION E-192 E-192, Standby Diesel Generator Loading Calculation

G13.18.2.7*22 Steam Tunnel leak Detection Temperature Limit Sensitivity to Initial Temperature and Leak Rate

G13.18.6.1-E31*009 Instrument Loop Uncertainty/Setpoint Determination for Main Steam Line Tunnel Ambient Temperature -

High 0

CONDITION REPORTS

CR-RBS-2004-01752

CR-RBS-2009-06462

CR-RBS-2007-02690

CR-RBS-2010-00910

CR-RBS-2009-06036

CR-RBS-2010-01981

CR-RBS-2009-06460

CR-RBS-2010-02468

Attachment

DRAWINGS

NUMBER TITLE REVISION 828E537AA Elementary Diagram HPCS Power Supply System 25

EE-001M 4160V One Line Diagram Standby Bus E22-S004 9

TLD-LMS-075 Test Loop Diagram Main Steam Line Tunnel Ambient Temperature E31-T/CN031A

GENERIC LETTER

NUMBER TITLE DATE

GL 79-36 Adequacy of Station Electric Distribution Systems Voltages August 8, 1979

INFORMATION NOTICES

NUMBER TITLE DATE NRC

IN 84-69 Operation of Emergency Diesel Generators August 29, 1984

NRC

IN 84-69, Supplement 1 Operation of Emergency Diesel Generators February 24, 1986 NCR

IN 91-29, Supplement 3 Deficiencies Identified During Electrical Distribution System Functional Inspections November 22, 1995

MISCELLANEOUS

NUMBER TITLE REVISION/DATE

IEEE Std 308-1974 IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations

1974 Vendor Technical Document No.

VTD-B455-0150 Brown Boveri (Now ABB Power T&D Company) Instructions for Solid-State Directional Relays [PUB. #IB 18.8.7-5D

Reactor Core Isolation Cooling Pump Discharge Line Loop Calibration Report Terry Steam Turbine Manual Section 9 6-79

Attachment

PROCEDURES

NUMBER TITLE REVISION

ARP-601-19 P601-19 Alarm Response 026

ARP-601-21 E-04, RCIC Turbine Bearing Oil Pressure Low 303

EOP-0003 Emergency Operating Procedure - Secondary Containment and Radioactive Release Control

014

EN-LI-102 Corrective Action Process 14

EN-OP-104 Operability Determinations 4

EN-OP-115 Conduct of Operations 9

STP-209-6310 RCIC Quarterly Pump and Valve Operability Test 32

TECHNICAL SPECIFICATIONS

NUMBER TITLE AMENDMENT

Technical Specification 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation

165 Technical Specification 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 81 Technical Specification 3.7.1 Standby Service Water (SSW) System and Ultimate Heat Sink (UHS)

Technical Specification 3.8.1 AC Sources - Operating 156 Technical Specification 3.8.9 Distribution Systems - Operating 81

UPDATED SAFETY ANALYSIS REPORT DOCUMENT

NUMBER TITLE REVISION USAR 10.3 Main Steam Supply System 14

Attachment

Section 1R18: Plant Modifications

PROCEDURES

NUMBER TITLE REVISION

STP-207-4201 NSSS Isolation-Main Steam Tunnel Temperature High Channel Calibration and LSFF (E31-N604A, E31-N604E, E31-R617E)

STP-207-4202 NSSS Isolation-Main Steam Tunnel Temperature High Channel Calibration and LSFF (E31-N604B, E31-N604F, E31-R617F)

STP-207-4203 NSSS Isolation-Main Steam Tunnel Temperature High Channel Calibration and LSFF (E31-N604C)

2

STP-207-4204 NSSS Isolation-Main Steam Tunnel Temperature High Channel Calibration and LSFF (E31-N604D)

303

TP-10-0001 Noblechem Application Procedure 00

WORK ORDER

NUMBER TITLE DATE

WO 00235801-01 Set Up Group I Isolation, Install

EC 21957 May 21, 2010

Section 1R19: Postmaintenance Testing

BULLETIN

NUMBER TITLE DATE IE Bulletin 85-03 Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings November 15, 1985

CONDITION REPORTS

CR-RBS-1999-01425

CR-RBS-2010-01237

CR-RBS-2010-02821

CR-RBS-2007-03506

CR-RBS-2010-02099

CR-RBS-2010-00556

CR-RBS-2010-02500

CR-RBS-2010-01059

CR-RBS-2010-02792

Attachment

DRAWING

NUMBER TITLE REVISION

ESK-06HVR10 Elementary Diagram 480V SWGR Containment Unit Cooler *UC1B

INFORMATION NOTICES

NUMBER TITLE DATE

IN 87-40 Backseating Valves Routinely to Prevent Packing Leakage August 31, 1987

IN 96-48, Supplement 1 Motor-Operated Valve Performance Issues July 24, 1998

PROCEDURES

NUMBER TITLE REVISION/DATE

EN-DC-304 MOV Thrust / Torque Setpoint Calculations 1 NRC

IP 62710 Power-Operated Gate Valve Pressure Locking and Thermal Binding January 17, 2001

OSP-0102 Turbine Valve Testing 305

STP-256-0203 Division I Cooling Tower Fans Operability Test 004

TECHNICAL SPECIFICATIONS

NUMBER TITLE AMENDMENT/REVISIONTechnical Specification 3.3.1.1 Reactor Protection System (RPS) Instrumentation Technical Specification 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation

Attachment

UPDATED SAFETY ANALYSIS REPORT DOCUMENTS

NUMBER TITLE DATE USAR 10.2 Turbine Generator August 1987

WORK ORDER

NUMBER TITLE DATE

WO 00225094-01

HVR-UC1B Failed to Start; Implement

MA-125 March 19, 2010

WO 00228482-01

MSS-HYVCV2 Fast Acting Solenoid Did not Actuate; IC Replace Fast Acting solenoid

TMB-SOV4FOB201SD1 March 11, 2010

WO 00238997-01 Electrically Backseat E51-MOVF064 IAW

EC-22647 Requirements May 28, 2010

Section 1R22: Surveillance Testing

CONDITION REPORTS

CR-RBS-2005-02124

CR-RBS-2007-05586

CR-RBS-2008-02751

CR-RBS-2010-02500

PROCEDURES

NUMBER TITLE REVISION

EN-OP-109

STP-256-0203 Drywell Leakage Division I Cooling Tower Fans Operability Test

004

Section 1EP6: Drill Evaluation

SCENARIO

NUMBER TITLE REVISION

RSMS-OPS-635 APRM F, Feed Pump Minimum Flow Valve, RWCU Leak - ATWS 0

Attachment

Section 1EP7: Force-on-Force Exercise Evaluation

PROCEDURES

NUMBER TITLE REVISION

AOP-0054

EIP-2-001 Security Events Classification of Emergencies

19 Section 2RS04:

Occupational Dose Assessment

PROCEDURES

NUMBER TITLE REVISION

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

EN-RP-104 Personnel Contamination Events 4

EN-RP-105 Radiation Work Permits 8

EN-RP-108 Radiation Protection Posting 8

EN-RP-201 Dosimetry Administration 3

EN-RP-202 Personnel Monitoring 7

EN-RP-203 Dose Assessment 3

EN-RP-204 Special Monitoring Requirements 3

EN-RP-205 Prenatal Monitoring 3

EN-RP-206 Dosimeter of Legal Record Quality Assurance 2

EN-RP-207 Planned Special Exposure 3

EN-RP-208 Whole Body Counting and In-Vitro Bioassay 3

EN-RP-311 Electronic Alarming Dosimeter 9

CONDITION REPORTS

CR-RBS-2009-01568

CR-RBS-2009-03953

CR-RBS-2009-04785

CR-RBS-2009-04786

CR-RBS-2009-04858

CR-RBS-2009-05239

CR-RBS-2009-05772

CR-RBS-2009-06017

CR-RBS-2010-01254

CR-RBS-2010-02058

CR-RBS-2010-02491

RADIATION WORK PERMITS

PROCEDURES

NUMBER TITLE

20091012 General FIN Team activities

20091434 RF15 LLRTs, including support activities

20091936 RF15 Installation/removal of temporary shielding

20091408 RF15 Operations Tagging Activities in LHRA except Drywell

20101011 Declared Pregnant Women authorized in RCA, RA

Attachment AUDITS,

SELF-ASSESSMENTS, AND SURVEILLANCES

NUMBER TITLE DATE

LO-RLO-2009-00143 Focused Assessment Report; Occupational Rad Safety January 22, 2010LO-RLO-2009-00057 Focused Assessment Report; Occupational Rad Safety

August 7, 2009 RP Snapshot Assessment July 20-22, 2009

MISCELLANEOUS DOCUMENT

TITLE DATE Whole Body Count Assessments and Records September 2009 - May 2010

Section 2RS05:

Radiation Monitoring Instrumentation

PROCEDURES

NUMBER TITLE REVISION

EN-RP-301 Radiation Protection Instrument Control 2

EN-RP-302 Operation of Radiation Protection Instrumentation 1

EN-RP-303 Source Checking of Radiation Protection Instrumentation 2

EN-RP-306 Calibration and Operation of the Eberline

PM-7 2

EN-RP-307 Operation and Calibration of the Eberline Personnel Contamination Monitors

EN-RP-308 Operation and Calibration of Gamma Scintillation Tool Monitors 2

EN-MA-105 Control of Measuring and Test Equipment 3

MCP-4201 DRMS Low Range Area Monitor Calibration 5

RHP-0106 Calibration of the Canberra Fastscan and Accuscan II Whole Body Counters

Attachment NUMBER TITLE REVISION

RPP-0036 Calibration of DRMS Area Monitors and Determination of Alert and High Alarm Setpoints

5A

STP-511-4201 Main Steam Line Radiation High High Calibration and Logic System Functional Test

STP-511-4203 Main Steam Line Radiation High High Calibration and Logic System Functional Test

AUDITS AND SELF-ASSESSMENTS

NUMBER TITLE DATE

QA 14-15-2009 Calibration Laboratory Audit Report and Central Calibration Facility Audit Checklist December 5, 2009

RADIATION PROTECTION INSTRUMENTATION CALIBRATIONS

Identification-Model No. Instrument Type Calibration Date Fastcan II

Accuscan II A006 #000201 A006 #000201 89-0560

PCM-1B #356

PCM-2 #614

21 436 Whole Body Counter Whole Body Counter Tool Monitor Tool Monitor Tool Monitor Personnel Contamination Monitor Personnel Contamination Monitor

PM-7

PM-7 March 4, 2010 March 2, 2010 September 1, 2009April 8, 2010 August 28, 2009 June 29, 2009

February 17, 2009 June 4, 2009 August 28, 2009

PROCESS EFFLUENT AND AREA MONITOR FUNCTIONAL TEST AND CALIBRATIONS

Effluent Monitor Calibrations Channel No. Monitor Description Procedure Calibration Dates

RE125

RE3A RE3A RE5A RE125 RE107 Plant vent exhaust noble gas Main Steam Line Monitor Main Steam Line Monitor Fuel Building Vent Exhaust Plant Vent Stack Flow Rate Liquid Radwaste Effluent Monitor

STP-511-4214

STP-511-4201

STP-511-4201

STP-511-4205

STP-511-4231

STP-511-4280 December 11, 2008 April 6, 2009 April 6, 2009 September 12, 2008 August 27, 2009 March 11, 2009

Attachment Area Monitor Calibrations Channel No.

Monitor Description Surveillance

Procedure

Calibration Dates

RE16A Primary Containment Area Radiation Monitor

STP-511-4249 March 9, 2010

RE16B

RE20A

RE145

RE143

Primary Containment Area Radiation Monitor Primary Drywell Area Radiation Monitor

Reactor Building Fuel Transfer Area

Reactor Water Cleanup Precoat Area

STP-511-4250

STP-511-4289

MCP-4201

MCP-4201 December 22, 2009

October 15, 2009

April 28, 2008

March 19, 2008

CONDITION REPORTS

CR-RBS-2009-01454

CR-RBS-2009-01501

CR-RBS-2009-01722 CR-RBS-2009-01830

Section 4OA1: Performance Indicator Verification

PROCEDURES

NUMBER TITLE REVISION

EN-LI-102 Corrective Action Process 15

EN-LI-114 Performance Indicator Process

4

Section 4OA2: Identification and Resolution of Problems

CONDITION REPORTS

CR-RBS-2001-01676

CR-RBS-2009-05590

CR-RBS-2010-00077

CR-RBS-2010-00526

CR-RBS-2010-02342

CR-RBS-2007-04490

CR-RBS-2009-05906

CR-RBS-2010-0026

CR-RBS-2010-00529

CR-RBS-2010-02466

CR-RBS-2009-05446

CR-RBS-2009-06132

CR-RBS-2010-00268

CR-RBS-2010-01854

CR-RBS-2009-05491

CR-RBS-2009-06148

CR-RBS-2010-00503

CR-RBS-2010-02099

Attachment MISCELLANEOUS

NUMBER TITLE REVISION A09462-R-001 DIV I SDG Intercooler Adapter Flange Bolt Failure Assessment

A09462-R-002 Structural Assessment of DIV I SDG Intercooler Adapter Flange with Additional Bolt Failures

F09496-R-001 RBS DIV I & II SDG Cyclic Load Investigation Test Data Report 0 F09496-R-002 RBS DIV I & II SDG Air Intake System Cyclic Load Investigation

F09496-R-003 RBS DIV I & II SDG Air Intake System Analysis and Load Investigation Summary and Recommendations Operational Decision Making Instruction Steam Leak in the 95' Steam Tunnel Main Steam Line Area 0

WORK ORDERS

00138862

00155267

00154836

00155268

00154838 00154840