IR 05000458/2016001

NRC Integrated Inspection Report 05000458/2016001 and Final Significance Determination of Green Finding; NRC Inspection Report 05000458/2016008

ML16132A144
Person / Time
Site:	River Bend
Issue date:	05/11/2016
From:	Greg Warnick NRC/RGN-IV/DRP/RPB-C
To:	Maguire W Entergy Operations
Warnick G
References
EA-13-110, EA-13-15-140 IR 2016001, IR 2016008
Download: ML16132A144 (50)

Text

May 11, 2016

SUBJECT:

RIVER BEND STATION - NRC INTEGRATED INSPECTION REPORT 05000458/2016001 AND FINAL SIGNIFICANCE DETERMINATION OF GREEN FINDING; NRC INSPECTION REPORT 05000458/2016008

Dear Mr. Maguire:

On March 31, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your River Bend Station, Unit 1. On April 14, 2016, the NRC inspectors discussed the results of this inspection with Mr. C. Rich, General Manager, Plant Operations, and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

This letter also provides you the final significance determination of the preliminary Greater than Green finding identified in NRC Inspection Report 05000458/2015010 (ML16047A268), dated February 16, 2016. A detailed description of the finding is contained in Section 2.6.a of that report. The finding was associated with the failure to adequately assess the increase in risk of operating the control building chilled water system chillers in various single-failure vulnerable configurations.

At your request, a Regulatory Conference was held on April 4, 2016, to discuss your position on the preliminary Greater than Green finding and to present new information based on testing conducted by your staff. A copy of your presentation provided at the Regulatory Conference is attached to the summary of the Regulatory Conference (ML16119A342), dated April 27, 2016.

In your presentation, you discussed information important to characterize the safety significance of the finding associated with the control building chilled water system. Specifically, you presented methodologies used by Entergy to determine realistic control room heat loads and develop an estimate of the control room heatup rate, taking into account control room habitability for operators and equipment design temperature limits.

After considering the information reviewed during our inspections and the information you provided at the Regulatory Conference, the NRC has concluded that the finding is appropriately characterized as Green, a finding of very low safety significance. See Section 4OA5 of this report for additional information. NRC inspectors documented one finding of very low safety significance (Green) in this report.

This finding involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement policy.

If you contest the violation or significance of this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the River Bend Station.

If you disagree with a cross-cutting aspect assignment 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 the River Bend Station.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management 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/

Gregory G. Warnick, Chief Project Branch C Division of Reactor Projects Docket No. 50-458 License No. NPF-47

Enclosure:

Inspection Report 05000458/2016001 w/ Attachments:

1. Supplemental Information 2. Request for Information for the O

REGION IV==

Docket: 05000458 License: NPF-47 Report: 05000458/2016001 Licensee: Entergy Operations, Inc.

Facility: River Bend Station Location: 5485 U.S. Highway 61N St. Francisville, LA 70775 Dates: January 1 through March 31, 2016 Inspectors: J. Sowa, Senior Resident Inspector S. Makor, Acting Senior Resident Inspector B. Parks, Acting Resident Inspector L. Brandt, Project Engineer J. Mateychick, Senior Reactor Inspector J. ODonnell, CHP, Health Physicist M. Phalen, Senior Health Physicist G. Guerra, CHP, Emergency Planning Inspector D. Bradley, Resident Inspector, Columbia Generating Station R. Deese, Senior Reactor Analyst Approved By: G. Warnick, Chief Project Branch C Division of Reactor Projects-1- Enclosure

SUMMARY

IR 05000458/2016001; 01/01/2016 - 03/31/2016; River Bend Station; Other Activities.

The inspection activities described in this report were performed between January 1 and March 31, 2016, by the resident inspectors at River Bend Station and inspectors from the NRCs Region IV office. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for the licensees failure to adequately assess the increase in risk that may result from proposed maintenance activities before performing maintenance activities. Specifically, prior to March 30, 2015, the risk assessment performed by the licensee for plant maintenance failed to account for certain safety significant structures, systems, and components that were concurrently out of service. On multiple occasions, the licensee failed to adequately assess the risk of operating the control building chilled water system (HVK) chillers in various single failure vulnerable configurations. As a result of this deficiency, the station reduced the reliability and availability of systems contained in the main control room and failed to account for the significant, uncompensated impairment of the safety functions of the associated systems. In response to the NRCs conclusions, the licensee initiated Condition Report CR-RBS-2016-00095. The licensee also completed engineering analyses to evaluate alternate cooling methods, including cross-connecting service water and the HVK chiller systems, in order to provide cooling to spaces housing electrical components and mitigate a loss of HVK event.

This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the licensees failure to account for a loss of all HVK cooling scenario, either quantitatively or qualitatively, resulted in uncompensated impairment to all systems associated within the main control room. A loss of cooling to the control room could lead to multiple systems exceeding their equipment qualification temperatures and impact control room habitability. The finding was evaluated using Inspection Manual Chapter (IMC) 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process.

Using Inspection Manual Chapter 0609, Appendix K, the finding was determined to require additional NRC management review using risk insights where possible because the quantitative probabilistic risk assessment (PRA) tools are not well suited to analyze failures from control room heat-up events.

A Region IV senior reactor analyst performed a detailed risk evaluation. This evaluation yielded a maximum incremental core damage probability deficit of 3.2E-7. The analyst applied this result to Flowchart 1, Assessment of Risk Deficit, of Appendix K. In applying Flowchart 1, the analyst determined that because the maximum incremental core damage probability deficit was less than 1.0E-6, the finding was of very low safety significance (Green).

The team determined the most significant contributing cause of the licensee failing to adequately assess the increase in risk from proposed maintenance activities was inadequate procedural guidance in Procedure ADM-0096, Risk Management Program Implementation and On-line Maintenance Risk Assessment, Revision 316. This finding has a resources cross-cutting aspect within the human performance area because leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety [H.1]. (Section 4OA5.2)

PLANT STATUS

River Bend Station began the inspection period at 100 percent reactor thermal power. It departed from full power as follows:

• On January 9, 2016, a reactor scram from 100 percent power occurred due to an electrical transient experienced during a severe thunderstorm. The station returned the unit to 100 percent power on February 7, 2016.

• On February 13, 2016, the station reduced power to 20 percent to change in-service steam jet air ejectors. The station returned the unit to 100 percent power on February 15, 2016.

• On February 17, 2016, the station conducted a shutdown in order to repair a crack in a joint between a drain line and the main condenser which impacted the ability to maintain condenser vacuum. The station returned the unit to 100 percent power on March 7, 2016.

Power remained at or near 100 percent for the remainder of the inspection period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On January 21, 2016, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensees procedures to respond to tornadoes and high winds, and the licensees planned implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constitute one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

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

• March 10, 2016, Division I residual heat removal system

• March 11, 2016, Division I standby service water

• March 28, 2016, Division II 125 volt dc power system The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems were correctly aligned for the existing plant configuration.

These activities constitute three partial system walkdown samples, as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On March 10, 2016, the inspectors performed a complete system walkdown inspection of the standby liquid control system. The inspectors reviewed the licensees procedures and system design information to determine the correct system lineup for the existing plant configuration. The inspectors also reviewed outstanding work orders, open condition reports, in-process design changes, temporary modifications, and other open items tracked by the licensees operations and engineering departments. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.

These activities constitute one complete system walkdown sample, as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on six plant areas important to safety:

• March 1, 2016, auxiliary building 141 elevation, fire area AB-15/Z-4

• March 1, 2016, auxiliary building 141 elevation, fire area AB-1/Z-4

• March 10, 2016, auxiliary building, residual heat removal pump room A, fire area AB-5

• March 11, 2016, containment building, standby liquid control system area, fire area RC-4/Z-4

• March 11, 2016, standby cooling tower pump room A, fire area PH-1/Z-1

• March 11, 2016, standby cooling tower pump room B, fire area PH-2/Z-1 For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constitute six quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

On March 13, 2016, the inspectors completed an inspection of the stations ability to mitigate flooding due to internal causes. After reviewing the licensees flooding analysis, the inspectors chose one plant area containing risk-significant structures, systems, and components (SSCs) that was susceptible to flooding:

• Division I residual heat removal pump room AB-070-2 The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.

In addition, on March 15, 2016, the inspectors completed an inspection of underground bunkers susceptible to flooding. The inspectors selected two underground electrical manholes that contained risk-significant or multiple-train cables whose failure could disable risk-significant equipment:

• Electrical manhole 1EMH101

• Electrical manhole 1EMH613 The inspectors observed the material condition of the cables and splices contained in the electrical manholes and looked for evidence of cable degradation due to water intrusion. The inspectors verified that the cables and vaults met design requirements.

These activities constitute completion of one flood protection measures sample and one bunker/manhole sample, as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On March 10, 2016, the inspectors observed a portion of an annual requalification test for licensed operators. The inspectors assessed the performance of the operators and the evaluators critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the requalification activities.

These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

On January 30, 2016, the inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity due to conducting a reactor startup following a forced outage.

In addition, the inspectors assessed the operators adherence to plant procedures, including the conduct of operations procedure and other operations department policies.

These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed two instances of degraded performance or condition of safety-related SSCs:

• March 23, 2016, standby switchgear air handling unit 2B, functional failure review

• March 30, 2016, reactor protection system, functional failure review The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constitute completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed four risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:

• January 5, 2016, green risk condition while residual heat removal system A was out of service for maintenance concurrent with the low pressure core spray pump out of service

• February 3, 2016, green risk condition following a failed surveillance test on control room air conditioning fan HVC-ACU1A

• March 16, 2016, green risk condition while residual heat removal system B was out of service for maintenance concurrent with standby gas treatment B out of service

• March 21, 2016, yellow risk condition while the Division III diesel generator was out of service for maintenance The inspectors verified that these risk assessments were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensees risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.

Additionally, on March 3, 2016, the inspectors observed portions of one emergent work activity, work on trip unit B-21ESN670E requiring disabling Division I alternate depressurization system (ADS) actuation logic that had the potential to affect the functional capability of mitigating systems.

The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected SSCs.

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

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed four operability determinations that the licensee performed for degraded or nonconforming SSCs:

• January 6, 2016, operability determination of agastat relays that provide trip signals to the reactor feedwater pumps and turbine on a reactor vessel high level condition (CR-RBS-2015-09109)

• March 14, 2016, operability determination of high auxiliary building pressure indications (CR-RBS-2016-00065)

• March 22, 2016, operability determination of lowering scram valve current indications during diesel generator operations (CR-RBS-2016-00024)

• March 22, 2016, operability determination of battery room 1B fans and dampers (CR-RBS-2014-03221)

The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC.

The inspectors reviewed operator actions taken or planned to compensate for degraded or nonconforming conditions. The inspectors verified that the licensee effectively managed these operator workarounds to prevent adverse effects on the function of mitigating systems and to minimize their impact on the operators ability to implement abnormal and emergency operating procedures.

These activities constitute completion of five operability and functionality review samples, which included one operator work-around sample, as defined in Inspection Procedure 71111.15.

b. Findings

No findings were identified.

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed two permanent plant modifications that affected risk-significant SSCs:

• February 22, 2016, motor generator A voltage regulator modification

• February 22, 2016, motor generator B voltage regulator modification The inspectors reviewed the design and implementation of the modifications. The inspectors verified that work activities involved in implementing the modifications did not adversely impact operator actions that may be required in response to an emergency or other unplanned event. The inspectors verified that post-modification testing was adequate to establish the functionality of the SSCs as modified.

These activities constitute completion of two samples of permanent modifications, as defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed seven post-maintenance testing activities that affected risk-significant SSCs:

• January 25, 2016, work order (WO) 00436165, Troubleshoot Div III Voltage Regulator, following Reactor Protection System (RPS) motor generator (MG)modification

• January 28, 2016, WO 00435379, Spurious E-40 Trip Code on Main Steam Tunnel High Ambient Temperature. Replace E31-N604D, following high temperature indication in main steam tunnel

• January 30, 2016, WO 00436558, C71-S003A Would Not Close When Placing MG Set In Service, following RPS motor generator modification

• February 24, 2016, WO 00438672, C51-JEN002D Troubleshoot Indication CR16-1612, following replacement of intermediate range monitor D detector driven gear box and limit switch adjustments

• February 29, 2016, WO 00438973, EJS-SWG2A ACB36 Remove the Close Signal Upon Breaker Closure, following identified issues potentially affecting Masterpact circuit breakers

• March 1, 2016, WO 00438978, EJS-SWG2B ACB65 Remove the Close Signal Upon Breaker Closure, following identified issues potentially affecting Masterpact circuit breakers

• March 8, 2016, WO 00439318, Replace Gear Box & Perform, Insert & Retract Test for IRM D, following intermediate range monitor D rod block The inspectors reviewed licensing- and design-basis documents for the SSCs and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected SSCs.

These activities constitute completion of seven post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

Following a reactor scram on January 9, 2016, the station placed shutdown cooling in service and cooled the plant down to Mode 4. The stations forced outage concluded on January 26, 2016.

Following a loss of condenser vacuum on February 17, 2016, the station conducted a shutdown, placed shutdown cooling in service, and cooled the plant down to Mode 4 for repairs. The stations forced outage concluded on March 4, 2016.

During the stations forced maintenance outages, the inspectors evaluated the licensees outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan and developed mitigation strategies for losses of key safety functions. This verification included the following:

• Monitoring of shutdown and cooldown activities

• Verification that the licensee maintained defense-in-depth during outage activities

• Monitoring of heat-up and start-up activities

• Observation and review of fuel handling activities These activities constitute completion of two outage activities samples, as defined in Inspection Procedure 71111.20.

b. Findings

In response to a loss of shutdown cooling that occurred on January 10, 2016, following an automatic reactor scram on January 9, 2016, a Special Inspection was performed by the NRC. Issues of concern associated with these activities were incorporated into the charter for the Special Inspection team that was onsite on February 8, 2016, and were dispositioned during the course of their inspection activities.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed five risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the SSCs were capable of performing their safety functions:

Reactor coolant system leak detection tests:

• January 27, 2016, STP-050-0702, Refueling Outage Reactor Pressure Vessel Inservice Leakage Test Other surveillance tests:

• January 25, 2016, STP-309-0203, Division III Diesel Generator Operability Test

• January 30, 2016, STP-508-1700, Division I C71-S003A & C Electrical Protection Assembly Channel Functional Test

• February 2, 2016, STP-0514217, Main Steam Line Isolation Channel Calibration Test and Logic System Functional Test, performed on January 29, 2016

• March 12, 2016, STP-256-6304, Standby Service Water B Loop Quarterly Pump and Valve Operability Test The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

These activities constitute completion of five surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

Training Evolution Observation

a. Inspection Scope

On March 17, 2016, the inspectors observed simulator-based licensed operator requalification training that included implementation of the licensees emergency plan.

The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the evaluators and entered into the corrective action program for resolution.

These activities constitute completion of one training observation sample, as defined in Inspection Procedure 71114.06.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors performed this portion of the attachment as a post-outage review. During the inspection the inspectors interviewed licensee personnel, reviewed licensee documents, and evaluated licensee performance in the following areas:

• Radiological work planning, including work activities of exposure significance, and radiological work planning ALARA evaluations, initial and revised exposure estimates, and exposure mitigation requirements. The inspectors also verified that the licensees planning identified appropriate dose reduction techniques, reviewed any inconsistencies between intended and actual work activity doses, and determined if post-job (work activity) reviews were conducted to identify lessons learned.

• Verification of dose estimates and exposure tracking systems including the basis for exposure estimates, and measures to track, trend, and if necessary reduce occupational doses for ongoing work activities. The inspectors evaluated the licensees method for adjusting exposure estimates and reviewed the licensees evaluations of inconsistent or incongruent results from the licensees intended radiological outcomes.

• Problem identification and resolution for ALARA planning and controls. The inspectors reviewed audits, self-assessments, work-in-progress and post-job ALARA reviews, and corrective action program documents to verify problems were being identified and properly addressed for resolution.

These activities constitute completion of two of the five required samples of occupational ALARA planning and controls, as defined in Inspection Procedure 71124.02.

2RS4 Occupational Dose Assessment

a. Inspection Scope

The inspectors evaluated the accuracy and operability of the licensees personnel monitoring equipment, verified the accuracy and effectiveness of the licensees methods for determining total effective dose equivalent, and verified that the licensee was appropriately monitoring occupational dose. The inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:

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

• The technical competency and adequacy of the licensees internal dosimetry program

• Adequacy of the dosimetry program for special dosimetry situations such as declared pregnant workers, multiple dosimetry placement, and neutron dose assessment

• Audits, self-assessments, and corrective action documents related to dose assessment since the last inspection These activities constitute completion of five samples of occupational dose assessment, as defined in Inspection Procedure 71124.04.

b. Findings

No findings were identified.

OTHER ACTIVITIES

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

4OA1 Performance Indicator Verification

.1 Unplanned Scrams per 7000 Critical Hours (IE01)

a. Inspection Scope

The inspectors reviewed licensee event reports for the period of January 1, 2015, through December 31, 2015, to determine the number of scrams that occurred. The inspectors compared the number of scrams reported in these licensee event reports to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constitute verification of the unplanned scrams per 7000 critical hours performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Unplanned Power Changes per 7000 Critical Hours (IE03)

a. Inspection Scope

The inspectors reviewed operating logs, corrective action program records, and monthly operating reports for the period of January 1, 2015, through December 31, 2015, to determine the number of unplanned power changes that occurred. The inspectors compared the number of unplanned power changes documented to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constitute verification of the unplanned power changes per 7000 critical hours performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Unplanned Scrams with Complications (IE04)

a. Inspection Scope

The inspectors reviewed the licensees basis for including or excluding in this performance indicator each scram that occurred from January 1, 2015, through December 31, 2015. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constitute verification of the unplanned scrams with complications performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Annual Follow-up of Selected Issues

a. Inspection Scope

The inspectors selected one issue for in-depth follow-up:

• On January 9, 2016, a lightning induced voltage transient resulted in a phase to phase fault on the line from Big Cajun to Fancy Point Switchyard and caused a reactor scram. As a result of this scram, the inspectors performed an in-depth assessment of the stations switchyard and sensitive equipment controls in order to assess the fault mechanism and evaluate the response of the reactor protection system. Specifically, the inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews, and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition. The inspectors discussed with licensee personnel and reviewed established administrative controls, methods for reviewing and scheduling activities, and responsibilities for maintenance and operation of the switchyard and transformer yard.

These activities constitute completion of one annual follow-up sample, as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

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

(Closed) Licensee Event Report (LER) 05000458/2015-004-00: Potential Loss of Safety Function of Onsite AC/DC Distribution Systems Due to Postulated Main Control Building Heat-up Following Loss of Ventilation Cooling System This LER describes the discovery of design basis calculation errors associated with the heat-up of certain building areas following the postulated failure of the heating, ventilation, and cooling (HVAC) system. The error failed to account for the additional heat load generated by upgraded AC/DC inverters installed in 2009. These calculation errors were reviewed during a special inspection that examined impacts on control room habitability and equipment functionality following a postulated loss of control building cooling. The results of this inspection are contained in NRC Inspection Report 05000458/2015010 (ADAMS ML16047A268). LER 05000458/2015-004-00 is closed.

These activities constitute completion of one event follow-up sample, as defined in Inspection Procedure 71153.

4OA5 Other Activities

.1 VIO 05000458/2013007-01, Failure to Resolve Noncompliances Associated with

Multiple Spurious Operations in a Timely Manner (EA-13-110) (Closed)

During an NRC inspection completed on December 30, 2013, the team identified a Green violation of License Condition 2.C.(10) for the failure to implement and maintain in effect all provisions of the approved fire protection program associated with multiple spurious operations concerns. Specifically, the licensee failed to implement all of the required corrective actions for multiple spurious operations concerns prior to November 2, 2012, which marked the expiration of enforcement discretion for multiple spurious operations contained in Enforcement Guidance Memorandum 09-002. Three scenarios were identified that negatively impacted the credited safe shutdown pumps, which take suction from the suppression pool and thereby reduced the volume of water available for safe shutdown.

The licensee performed Calculation G13.18.12.2-143, MSO [Multiple Spurious Operations] Scenario LPCI-A-4: Estimating Upper Pool Volume and Evaluating ECCS Pumps for Adequate NPSH, Revision 0, which confirmed that this multiple spurious operations scenario would not adversely impact the plants ability to achieve post-fire safe shutdown. The licensee performed Calculation G13.18.12.2-142, MSO Scenario RCIC-2: Estimating Percent of Suction Flow From CST Versus Suppression Pool, Revision 0, which confirmed that this multiple spurious operations scenario would not adversely impact the plants ability to achieve post-fire safe shutdown. The licensee evaluated the multiple spurious operations scenario involving spurious operation of a residual heat removal pump without a minimum flow discharge path. The licensee took actions to establish the time available for operators to identify and respond to the spurious operation before increased pump seal leakage would be a concern; confirmed the fire areas in the plant where the scenario was possible; and incorporated appropriate guidance for the control room operators in Procedure AOP-0052, Fire Outside The Main Control Room In Areas Containing Safety Related Equipment, Revision 25. The inspectors review confirmed that the licensees actions were appropriate and complete.

Therefore, this violation is closed.

.2 Failure to Adequately Assess Risk During Chiller Unavailability

a. Inspection Scope

This finding was documented in NRC Inspection Report 05000458/2015010 (AV 05000458/2015010-02, Section 2.6.a) as an apparent violation with potential Greater than Green significance (EA-15-140). The Special Inspection Team (team) reviewed the licensees corrective action documents, design calculations, and external engineering audit results. Additionally, the team reviewed control room heat load test results, performed in November 2015 and February 2016, by the licensee to demonstrate realistic heat loads. The team also reviewed the licensees procedures, testing methodology, sensitivity analysis, and conclusions to better inform the risk evaluation of a control room heat-up scenario, such as during a loss of all control building chilled water system (HVK) chillers event.

b. Findings

Introduction.

The inspectors identified a Green, non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for the licensees failure to adequately assess the increase in risk that may result from proposed maintenance activities before performing maintenance activities. Specifically, prior to March 30, 2015, the risk assessment performed by the licensee for plant maintenance failed to account for certain safety significant structures, systems, and components that were concurrently out of service.

On multiple occasions, the licensee failed to adequately assess the risk of operating the control building chilled water system (HVK) chillers in various single failure vulnerable configurations. As a result of this deficiency, the station reduced the reliability and availability of systems contained in the main control room and failed to account for the significant, uncompensated impairment of the safety functions of the associated systems.

Description.

The team reviewed the operational history of the HVK system and the licensees actions related to implementation of technical specifications for various HVK system configurations. This chilled water system supplies water during normal, shutdown, and design basis accident (DBA) conditions to the cooling coils in the main control room air conditioning units, standby switchgear rooms air conditioning units, and chiller equipment rooms air conditioning units. Chilled water is supplied by two independent trains, either one of which is capable of meeting the total chilled water demand. Each train contains two 100-percent capacity electric motor-driven, centrifugal liquid chillers (HVK chillers), two 100-percent capacity chilled water recirculation pumps, two 100-percent capacity condenser cooling water pumps, and one chilled water compression tank. The service water systems provide the chiller condenser cooling water. Any one running HVK chiller is sufficient to meet 100 percent of the total chilled water demand. The division I HVK chillers are labeled 1A and 1C. The division II HVK chillers are labeled 1B and 1D. The station typically operates with only one chiller running and the other three in a standby status.

The team noted that when an entire division of HVK chillers is out of service, such as chillers 1A and 1C for division I, the licensee would only enter the Technical Specification (TS) 3.7.3, Control Room Air Conditioning (AC) System, action statement for the condition of one control room AC subsystem being inoperable (condition A). The licensee did not enter TS action statements associated with inoperability of other components cooled by HVK chillers, such as the AC switchgear, DC switchgear, and vital inverters. An unresolved item, URI 05000458/2015010-01, Technical Specification Allowed Outage Time During Loss of Non-Technical Specification Supported Systems, was opened to resolve TS questions associated with the HVK system. Further discussion of the URI is included in Section 2.2.b of NRC inspection report 05000458/2015010.

The team then reviewed the licensees practices for assessing and managing risk, when removing HVK chillers from service, per 10 CFR 50.65(a)(4) and as described in the bases for TS 3.0.6. The River Bend Station utilizes a quantitative, level-1 probabilistic safety analysis (PSA) computer model named, Equipment Out of Service Monitor (EOOS). Licensee procedure ADM-0096, Risk Management Program Implementation and On-line Maintenance Risk Assessment, Revision 316, implements the requirements of 10 CFR 50.65(a)(4) and provides guidance on how and when to perform risk assessments using quantitative and qualitative tools.

Section 5.3 of procedure ADM-0096, Risk Assessment Overview, states the following regarding use of the EOOS computer model:

The Risk Assessment Program is a Risk-Informed Program, not a Risk Tool Based Program. This means that the quantitative results provided by the EOOS software must be blended with the qualitative guidance, in order to provide a complete risk picture of the situation. Decisions should never be made based on the EOOS quantitative results aloneQualitative factors (such as industry operating experience, personnel judgment, etc.) must also be used for fully assessing the effects of equipment out of service on plant risk.

The team noted that HVK chillers were modeled in EOOS and that the licensee would, using the computer program, disable the affected HVK chillers for a given maintenance period to yield a quantitative risk value.

The team then assessed the application of procedure ADM-0096 to specific work periods where multiple HVK chillers were removed from service simultaneously. For example, starting on December 15, 2014, the licensee removed HVK chillers 1A, 1B, and 1D from service for 41.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. During this work window, only the 1C chiller was available to provide cooling for both divisions of control room air conditioning, both divisions of AC switchgear, both divisions of DC switchgear, and both divisions of vital inverters. The licensee had assessed risk as 7.9 (Yellow) which included the quantitative tool (EOOS) and some qualitative factors for fire scenarios.

The team discovered that EOOS, however, did not model a control room heat-up scenario, such as during a loss of all HVK chillers event. The subsequent effects of failures of numerous control room components across multiple safety systems were, therefore, also not modeled due to the complexity of the event. The team reviewed procedure ADM-0096 for guidance on limitations of the PRA model and noted section 5.2.3 stated the following:

When the quantitative assessment tool is not available or the assessment scope is outside the scope of the EOOS risk monitor, qualitative assessments shall be performed.

The team also reviewed NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4A. Section 11.3.7.1 of NUMARC 93-01 discusses establishing action thresholds based on qualitative considerations.

This [qualitative] approach typically involves consideration of the following factors from the assessment:

• Duration of out-of-service condition, with longer duration resulting in increased exposure time to initiating events

• The number of remaining success paths (redundant systems, trains, operator actions, recovery actions) available to mitigate the initiating events The above factors can be used as the basis for establishment of a matrix or list of configurations and attendant risk management actions.

The team determined that the licensee did not consider the listed factors for qualitative assessments for a loss of control room cooling event. Specifically, the licensee did not establish a more limiting duration for placing control building chilled water system chillers in single-failure vulnerable configurations and, instead, relied upon the associated TS allowed outage time of 30 days. Further, the licensee did not consider the remaining success path, cross-connecting the HVK chillers with service water, and apply, as an example, just-in-time training or daily control room briefings on the procedure during single-failure vulnerable configurations with the potential to lose control room air conditioning.

The team noted that procedure ADM-0096 does not provide further guidance on how to qualitatively assess risk of a loss of all HVK cooling. Interestingly, attachment 7 of procedure ADM-0096 describes how to qualitatively assess and manage risk for removing non-TS auxiliary building cooling (HVR) due to EOOS limitations:

Auxiliary Building unit coolers HVR-UC11A(B) are non-Technical Specification equipment that provide cooling for safety-related equipment in the AB141 and AB170 locations. Each of these unit coolers are capable of providing the required cooling for both safety-related divisions. These unit coolers do not impact quantitative risk as determined using the EOOS risk monitor. To qualitatively address risk if HVR-UC11A(B) are unavailable, the following actions should be taken if one of HVR-UC11A(B) will be out of service[bulleted list of 13 actions].

Ultimately, the team determined that the licensee failed to adequately assess the risk of operating the control building chilled water system chillers in various single-failure vulnerable configurations. As a result of this deficiency, the station reduced the reliability and availability of systems contained in the main control room and failed to account for the significant, uncompensated impairment of the safety functions of the associated systems.

The licensee, in the example starting December 15, 2014, did not perform a qualitative risk assessment for a complete loss of control room cooling due to the inadequate procedural guidance in procedure ADM-0096. With an inadequate risk assessment of HVK system maintenance, the licensee did not appropriately determine the duration of the maintenance activity as described in the bases for TS 3.0.6.

To understand the exposure time for inadequate risk assessments, the team reviewed maintenance and TS data from control room logs. The team determined that the licensee operated in single-failure vulnerable configurations for the HVK system for 591 hours0.00684 days <br />0.164 hours <br />9.771825e-4 weeks <br />2.248755e-4 months <br /> over a 12 month period or approximately 6.7 percent of a year.

In response to the NRCs conclusions, the licensee initiated Condition Report CR-RBS-2016-00095. The licensee also contracted for an engineering analysis to credit alternate cooling methods, including cross-connecting service water and the HVK chiller systems, in order to cool vital electrical components and mitigate a loss of HVK event.

Analysis.

The team determined that the licensees failure to adequately assess the risk of operating the control building chilled water system chillers in various single-failure vulnerable configurations was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by not determining an appropriate duration of maintenance activities.

The team noted that, from section 7 of IMC 0612 Appendix E, Examples of Minor Issues, that discusses the Maintenance Rule, the performance deficiency is more than minor since the risk assessment failed to account for (at least qualitatively) the loss or significant, uncompensated impairment of a key operating or shutdown safety function.

Specifically, the licensees failure to account for a loss of all HVK cooling scenario, either quantitatively due to EOOS model limitations or qualitatively due to procedure inadequacies, represents a significant impairment to all systems associated with the main control room. A loss of cooling to the control room could lead to multiple systems exceeding their equipment qualification temperatures and lead to subsequent failures. A loss of cooling to the control room could also impact control room habitability. The team also reviewed NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4A. Section 11.3.7.1 of NUMARC 93-01 discusses establishing action thresholds based on qualitative considerations.

This [qualitative] approach typically involves consideration of the following factors from the assessment:

• Duration of out-of-service condition, with longer duration resulting in increased exposure time to initiating events

• The number of remaining success paths (redundant systems, trains, operator actions, recovery actions) available to mitigate the initiating events The above factors can be used as the basis for establishment of a matrix or list of configurations and attendant risk management actions.

The team determined that the licensee did not consider the listed factors for qualitative assessments for a loss of control room cooling event. This conclusion further supports the IMC 0612 Appendix E examples of more than minor performance deficiencies associated with the Maintenance Rule.

The finding was evaluated using Inspection Manual Chapter (IMC) 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Using Inspection Manual Chapter 0609, Appendix K, the finding required additional internal NRC management review using risk insights where possible because the quantitative probabilistic risk assessment (PRA) tools were not well suited to analyze failures from control room heat-up events. Thus, the analyst evaluated the safety significance posed by the heat-up of components cooled by control building chilled water (HVK) chillers using Appendix K, Flowchart 1, Assessment of Risk Deficit, to the extent practical, with additional risk insights by internal NRC management review in accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports. In accordance with Step 4.1.2 of Appendix K, the analyst performed a detailed risk evaluation for the Greater than Green Flowchart 1 result. The detailed risk evaluation resulted in a preliminary determination of White (low to moderate safety significance).

After considering information presented at the Regulatory Conference conducted April 4, 2016, a Region IV senior reactor analyst performed a final detailed risk evaluation. See 3 of this report, Final Detailed Risk Evaluation, for further information. This evaluation yielded a maximum incremental core damage probability deficit of 3.2E-7.

The analyst applied this result to Flowchart 1, Assessment of Risk Deficit, of Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, of Manual Chapter 0609. In applying Flowchart 1, the analyst determined that because the maximum incremental core damage probability deficit was less than 1.0E-6, the finding was of very low safety significance (Green).

The team determined the most significant contributing cause of the licensee failing to adequately assess the increase in risk from proposed maintenance activities involved inadequate procedural guidance in procedure ADM-0096. This finding has a resources cross-cutting aspect within the human performance area because leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety [H.1].

Enforcement.

10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), requires, in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance) the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, prior to March 30, 2015, before performing maintenance activities, the licensee failed to adequately assess the increase in risk that may result from proposed maintenance activities. Specifically, the risk assessment performed by the licensee for plant maintenance failed to account for certain safety significant structures, systems, and components that were concurrently out of service.

On multiple occasions, the licensee failed to adequately assess the risk of operating the control building chilled water system chillers in various single-failure vulnerable configurations. As a result of this deficiency, the station reduced the reliability and availability of systems contained in the main control room and failed to account for the significant, uncompensated impairment of the safety functions of the associated systems. In response to the NRCs conclusions, the licensee initiated Condition Report CR-RBS-2016-00095. The licensee also contracted for an engineering analysis to credit alternate cooling methods, including cross-connecting service water and the HVK chiller systems, in order to cool vital electrical components and mitigate a loss of a HVK event.

Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a. of the NRC Enforcement Policy: NCV 05000458/2015010-02, Failure to Adequately Assess Risk During Chiller Unavailability.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On January 15, 2016, the inspector presented the results associated with the closure of Violation 0500458/2013007-01 via phone to Mr. J. Clark, Manager, Regulatory Assurance.

The licensee acknowledged the issues presented.

On January 21, 2016, the inspectors presented the radiation safety inspection results to Mr. D. Burnett, Acting Director, Regulatory and Performance Improvement and Mr. S. Vazquez, Director, Engineering, and other members of the licensee staff. On April 27, 2016, the inspector presented the inspection results to Mr. J. Clark, Manager, Regulatory Assurance, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On April 14, 2016, the inspectors presented the inspection results to Mr. C. Rich, General Manager, Plant Operations, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On May 2, 2016, the Special Inspection Team conducted a telephonic exit to present the final significance determination results for AV 05000458/2015010-02, Failure to Adequately Assess Risk During Chiller Unavailability, to Mr. W. Maguire, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Burnett, Director, Emergency Planning, Entergy South

M. Chase, Director, Regulatory & Performance Improvement

J. Clark, Manager, Regulatory Assurance

B. Cole, Senior Manager, Fleet Radiation Protection

R. Conner, Manager, Nuclear Oversight

R. Cook, Manager, Security

K. Crissman, Senior Manager, Maintenance

L. Dautel, Supervisor, ALARA, Radiation Protection

D. Fletcher, Manager, Supply Chain

B. Ford, Senior Manager, Fleet Regulatory Assurance

T. Gates, Manager, Operations Support

J. Henderson, Manager, Systems & Components Engineering

B. Hite, Supervisor, Radiation Protection

K. Huffstatler, Senior Licensing Specialist, Regulatory Assurance

R. Leasure, Superintendent, Radiation Protection

P. Lucky, Manager, Performance Improvement

W. Maguire, Site Vice President

L. Meyer, Health Physics and Chemistry Specialist, Radiation Protection

C. Miller, Manager, Site Projects and Maintenance Services

J. Morgan, Senior HP/Chemistry Specialist, Chemistry

P. OConner, Manager, Training

S. Peterkin, Manager, Radiation Protection

J. Reynolds, Senior Manager, Operations

C. Rich, General Manager, Plant Operations

D. Sandlin, Manager, Design & Program Engineering

T. Schenk, Manager, Emergency Preparedness

S. Vazquez, Director, Engineering

T. Venable, Assistant Manager, Operations

J. Vukovics, Supervisor, Reactor Engineering

J. Wieging, Senior Manager, Production

J. Wilson, Manager, Chemistry

Attachment 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Closed

Potential Loss of Safety Function of Onsite AC/DC Distribution Systems Due to Postulated Main Control

05000458/2015-004-00 LER Building Heat-up Following Loss of Ventilation Cooling System (Section 4OA3)

Failure to Resolve Noncompliances Associated with

05000458/2013007-01 VIO Multiple Spurious Operations in a Timely Manner (Section 4OA5.1)

Failure to Adequately Assess Risk During Chiller

05000458/2015010-02 NCV Unavailability (Section 4OA5.2)