IR 05000382/2014003

IR 05000382-14-003; 04/01/2014 - 06/30/2014; Waterford Steam Electric Station, Unit 3; Integrated Inspection Report; Radiological Hazard Assessment and Exposure Controls and Problem Identification and Resolution

ML14223A545
Person / Time
Site:	Waterford
Issue date:	08/08/2014
From:	Greg Werner NRC/RGN-IV/DRP/RPB-E
To:	Chisum M Entergy Operations
References
IR-14-003
Download: ML14223A545 (60)

Text

ugust 8, 2014

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2014003

Dear Mr. Chisum:

On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Waterford Steam Electric Station, Unit 3. On July 1, 2014, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented two findings of very low safety significance (Green) in this report.

These findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these NCVs, 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 Waterford Steam Electric Station, Unit 3.

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 Waterford Steam Electric Station, Unit 3. 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 E. Werner, Chief Project Branch E Division of Reactor Projects Docket: 50-382 License: NPF-38

Enclosure:

Inspection Report 05000382/2014003 w/ Attachment: Supplemental Information

REGION IV==

Docket: 05000382 License: NPF-38 Report: 05000382/2014003 Licensee: Energy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: 17265 River Road Killona, LA 70057 Dates: April 1 through June 30, 2014 Inspectors: M. Davis, Senior Resident Inspector F. Ramirez, Senior Resident Inspector C. Speer, Resident Inspector J. Melfi, Project Engineer S. Hedger, Operations Inspector D. You, Project Engineer I. Anchondo, Senior Reactor Inspector L. Carson II, Senior Health Physicist P. Jayroe, Reactor Inspector J. ODonnell, Health Physicist R. Latta, Senior Reactor Inspector G. Guerra, CHP, Emergency Preparedness Inspector Approved Gregory E. Werner, Chief By: Project Branch E Division of Reactor Projects Enclosure

SUMMARY

IR 05000382/2014003; 04/01/2014 - 06/30/2014; Waterford Steam Electric Station,

Unit 3; Integrated Inspection Report; Radiological Hazard Assessment and Exposure Controls and Problem Identification and Resolution The inspection activities described in this report were performed between April 1 and June 30, 2014, by the resident inspectors at the site and inspectors from the NRCs Region IV office. Two findings of very low safety significance (Green) are documented in this report.

Both of these findings involved violations 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, Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Cornerstone: Occupational Radiation Safety

Green.

The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638.

The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes [H.12].

(Section 2RS1)

Cornerstone: Emergency Preparedness

Green.

The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10).

The licensee implemented compensatory measures while the system was being restored.

Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860.

The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609,

Attachment 4, Initial Characterization of Findings; and the corresponding Appendix B,

Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes [P.2]. (Section 4OA2.2)

Licensee-Identified Violations

None.

PLANT STATUS

The Waterford Steam Electric Station, Unit 3, began the inspection period at 100 percent power.

On April 12, 2014, operators commenced a down power to conduct activities associated with refueling outage 19. On May 12, 2014, operators commenced a reactor startup. Due to inspection and maintenance activities on feedwater pump B, reactor power was held at 49 percent from May 14, 2014, to May 25, 2014. On May 26, 2014, the unit reached 100 percent power and stayed there 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 May 28, 2014, the National Weather Service declared a tornado watch in the vicinity of the facility. The inspectors reviewed the plant personnels overall preparations and protection for the expected weather conditions. The inspectors evaluated the plant staffs preparations against the sites procedures and determined that the licensees actions were adequate. During the inspection, the inspectors focused on plant-specific design features and the licensees procedures used to respond to specified adverse weather conditions. The inspectors also toured the plant grounds to look for any loose debris that could become missiles during a tornado. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant. Additionally, the inspectors reviewed the final safety analysis report and performance requirements for the systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. The inspectors also reviewed a sample of corrective action program items to verify that the licensee-identified adverse weather issues were at an appropriate threshold and dispositioned them through the corrective action program in accordance with station corrective action procedures. Specific documents reviewed during this inspection are listed in the attachment.

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

b. Findings

No findings were identified.

1R04 Equipment Alignment

Partial Walkdown

a. Inspection Scope

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

• On April 29, 2014, auxiliary component cooling water train A during planned maintenance on auxiliary component cooling water train B

• On April 29, 2014, low pressure safety injection train A during planned maintenance on low pressure safety injection train B

• On April 29, 2014, spent fuel pool cooling train A during planned maintenance on spent fuel pool cooling train B

• On May 6, 2014, low pressure safety injection train B following an extended system outage 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 trains were correctly aligned for the existing plant configuration.

These activities constituted four partial system walk-down samples 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 five plant areas important to safety:

• On April 3, 2014, fire area CT1-001, dry cooling tower A

• On April 3, 2014, fire area CT3-001, wet cooling tower A

• On April 17, 2014, fire area RCB, reactor containment building

• On May 11, 2014, fire area RAB-007, relay room

• On May 11, 2014, fire area TB-005, turbine building operating floor +67.00 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 constituted five quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

On June 17, 2014, the inspectors completed an inspection of underground bunkers susceptible to flooding. The inspectors selected two underground bunkers that contained risk-significant or multiple-train cables whose failure could disable risk-significant equipment:

• Manhole M308-NA

• Manhole M317-NA The inspectors observed the material condition of the cables and splices contained in the bunkers 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 bunker/manhole sample, as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

On June 25, 2014, the inspectors completed an inspection of the readiness and availability of risk-significant heat exchangers. The inspectors verified the licensee used the industry standard periodic maintenance method outlined in Electric Power Research Institute NP-7552 for the train A shutdown cooling heat exchanger. Additionally, the inspectors walked down the heat exchanger to observe its performance and material condition and verified that the heat exchanger was correctly categorized under the Maintenance Rule and was receiving the required maintenance.

These activities constitute completion of one heat sink performance annual review sample, as defined in Inspection Procedure 71111.07.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

The activities described in subsections 1 through 4 below constitute completion of one inservice inspection sample, as defined in Inspection Procedure 71111.08.

.1 Non-destructive Examination Activities and Welding Activities

a. Inspection Scope

The inspectors directly observed portions of the following nondestructive examinations:

System Weld Identification Examination Type Safety Injection Reactor Cooling Loop 2B UT Safety Injection Piping Welds, Weld Numbers20-002 and 20-003 Main Feed Steam Generator #2 Main Feed Inlet UT Piping, Weld Number 46-004-XI Reactor Cooling Reactor Cooling Loop 2A and 2B Encoded Phased Safety Injection Dissimilar Metal Array UT Welds The inspectors reviewed records for the following nondestructive examinations:

System Weld Identification Examination Type Main Feed Main Feedwater 188B (Drain Valve) PT Assembly Safety Injection Safety Injection Replacement Valve RT SI512B Welds Reactor Cooling Reactor Coolant Pump 1A Inlet Encoded Phased Elbow to Safe-end Array UT During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the American Society of Mechanical Engineers Code requirements and applicable procedures. The inspectors also reviewed the qualifications of all nondestructive examination technicians performing the inspections to determine whether they were current.

The inspectors directly observed a portion of the following welding activities and reviewed the associated weld records:

System Weld Identification Examination Type Main Feed Main Feedwater 188B PT (Drain Valve) Assembly to Main Feedwater Piping Weld

b. Findings

No findings were identified.

.2 Vessel Upper Head Penetration Inspection Activities

Vessel upper head penetration inspection activities were not performed during Refuel 19. A new reactor head was installed in Refuel 18, bare metal visual inspections are scheduled for Refuel 21, and volumetric examinations are scheduled for Refuel 24.

.3 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensees boric acid corrosion control walk-down as specified in Procedure SEP-BAC-WF3-001, Boric Acid Corrosion Control Program, Revision 1.

The inspectors reviewed whether the visual inspections emphasized locations where boric acid leaks could cause degradation of safety significant components, and whether engineering evaluation used corrosion rates applicable to the affected components and properly assessed the effects of corrosion induced wastage on structural or pressure boundary integrity. The inspectors observed whether corrective actions taken were consistent with the American Society of Mechanical Engineers Code and 10 CFR 50, Appendix B, requirements.

b. Findings

No findings were identified.

.4 Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspectors reviewed the steam generator tube eddy current examination scope and expansion criteria to determine whether the criteria met technical specification requirements, Electric Power Research Institute guidelines, and commitments made to the NRC. The inspectors also reviewed whether the eddy current inspection scope included areas of degradations that were known to represent potential eddy current test challenges such as the top of tube sheet, tube support plates, and U-bends. The inspectors also observed the plugging of four tubes. The scope of the licensees eddy current examinations included:

• 100 percent full length bobbin coil inspection, Rows 3 and above

• 100 percent full length bobbin coil inspection, Rows 1 and 2 straight leg only

• 100 percent mid-range +Pt, Rows 1 and 2 U-bend from top of tube support sheet to top of support sheet

• Multiple +Pt inspections of special interest

• Channel head bowl visual inspection including divider plate juncture The inspectors reviewed the licensees identification of the following tube degradation mechanisms:

• Anti-vibration bar wear

• Tube support plate wear

• Loose part/foreign object wear The eddy current inspection constituted the first inspection after the replacement of both steam generators. The licensee did not identify any significant degradation of tubes; a total of four tubes were plugged, attributed to anti-vibration bar wear. All tubes plugged were on steam generator SG32 with one tube having a 25 percent through-wall indication, one tube with a 15 percent through-wall indication, and two tubes with 14 percent through-wall indication. The licensee will skip two cycles before the next steam generator eddy current inspection.

The inspectors observed portions of the eddy current testing being performed to determine whether

(1) the appropriate probes were used for identifying the expected types of degradation;

(2) calibration requirements were followed; and

(3) probe travel speed was in accordance with procedural requirements. The inspectors performed a review of the site-specific qualifications for the techniques being used to determine whether eddy current test data analyses were adequately performed per Electric Power Research Institute and site-specific guidelines.

The inspectors observed portions of secondary side feed ring inspection activities.

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 April 4, 2014, the inspectors observed a simulator training scenario prior to actual plant activities performed by an operating crew. 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 April 12, 2014, 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 shutdown activities associated with refueling outage 19. The inspectors observed the operators performance of the following activities:

• Plant shutdown, including the pre-job brief; and

• Reactor refueling activities In addition, the inspectors assessed the operators adherence to plant procedures, including 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 structures, systems, and components:

• On April 24, 2014, dry cooling tower train A component cooling water outlet header check valve (CC-181A)

• On May 10, 2014, containment fan cooler C component cooling water inlet isolation valve (CC-807A)

The inspectors reviewed the extent of condition of possible common cause of structure, system, and component 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 structures, systems, and components. 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 constituted 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 two 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:

• On May 1, 2014, scheduled maintenance activity on the train B component cooling water system with the B static uninterrupted power supply inverter being out-of-service

• On May 6, 2014, scheduled maintenance activity for the removal of the reactor vessel head with safety injection tanks being out-of-service 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.

The inspectors also observed portions of three emergent work activities that had the potential to cause an initiating event, to affect the functional capability of mitigating systems, or to impact barrier integrity:

• On April 22, 2014, emergent work on reactor coolant system hot leg temperature instrumentation

• On May 12, 2014, emergent maintenance activity on the main steam line B snubber

• On May 22, 2014, emergent maintenance activity on an emergency feedwater valve (EFW-228B) operator with the switchgear air-handling unit AH-30B being out-of-service 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 structures, systems, and components.

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 five operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components:

• On April 24, 2014, dry cooling tower fan 12B circuit breaker tripped

• On April 29, 2014, emergency diesel generator A fuel oil storage tank

• On April 30, 2014, auxiliary component cooling water header A to essential chillers isolation valve (ACC-112A)

• On May 12, 2014, emergency feedwater pump AB

• On June 17, 2014, auxiliary component cooling water header A to component cooling water heat exchanger outlet temperature control valve (ACC-126A)

The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded structures, systems, or components 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 structures, systems, or components.

These activities constitute completion of five operability and functionality review samples, 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 structures, systems, and components:

• On April 28, 2014, EC 44782, electronic governor replacement for emergency diesel generator A

• On April 29, 2014, EC 43812, modifications related to the feed water line 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 operability of the structures, systems, and components 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 structures, systems, or components:

• On April 25, 2014, testing following preventive maintenance activities on safety bus 31A

• On April 27, 2014, testing following permanent plant modification activities to install a digital governor on emergency diesel generator A

• On April 28, 2014, testing following preventive maintenance activities auxiliary component cooling water train A

• On April 28, 2014, adjusted stops on dry cooling tower train A component cooling water inlet isolation valve (CC-135A)

• On April 29, 2014, reactor coolant loop 2 shutdown cooling suction header isolation valve SI-407A

• On May 8, 2014, as left testing following corrective maintenance activities on electrical penetration 132

• On May 28, 2014, testing following preventive maintenance activities on train A emergency feedwater pump The inspectors reviewed licensing- and design-basis documents for the structures, systems, or components 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 structures, systems, or components.

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

On April 12, 2014, through May 11, 2014, during the stations refueling outage, the inspectors evaluated the licensees outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions. This verification included the following:

• Review of the licensees outage plan prior to the outage

• Monitoring of shut-down and cool-down activities

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

• Observation and review of reduced-inventory and mid-loop activities

• Observation and review of fuel handling activities

• Monitoring of heat-up and startup activities These activities constitute completion of one refueling outage sample as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed seven risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, and components were capable of performing their safety functions:

In-service tests:

• On April 9, 2014, containment spray pump A operability run Containment isolation valve surveillance tests:

• On April 21, 2014, letdown outside containment isolation valve CVC-109

• On April 21, 2014, fire protection header B outside containment isolation valve FP-601A Other surveillance tests:

• On April 10, 2014, fire protection system valve cycling

• On April 11, 2014, main steam safety valve simmer testing

• On April 18, 2014, high pressure safety injection AB flow balance

• On May 11, 2014, pressurizer heater emergency power supply functional test The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected structures, systems, and components following testing.

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

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

The inspectors assessed the licensees performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensees implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas:

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

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

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

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

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

• Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection These activities constitute completion of one sample of radiological hazard assessment and exposure controls as defined in Inspection Procedure 71124.01.

b. Findings

Introduction.

The inspectors reviewed a self-revealing, Green, non-cited violation of Technical Specification 6.12.1 that resulted when an individual entered a posted high radiation area without being on a high radiation area radiation work permit and was not knowledgeable of the dose rates.

Description.

On April 14, 2014, a security officer entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, without having knowledge of the actual radiological conditions. The individual was not on a radiation work permit that authorized entry into high radiation areas. According to licensee records, the individual entered the posted high radiation area twice and received an electronic dosimeter rate alarm of 107 millirem per hour. The individual observed the posted high radiation area and assumed that he was allowed to patrol the area. After hearing the electronic dosimeter alarm, the individual exited the room and when the electronic dosimeter stopped alarming, the guard re-entered the room to proceed with his patrol. When the individual heard the electronic dosimeter alarm, he did not stop and notify radiation protection of the electronic dosimeter alarm. When the individual exited the radiologically controlled area, he did not inform radiation protection of his electronic dosimeter alarm or that the radiologically controlled area access system had given him a red screen, whereby he was given another opportunity to inform radiation protection.

When the individual tried to enter the radiologically controlled area later in the shift, the radiologically controlled area access system would not allow his entry and directed him to the attention of radiation protection.

Analysis.

The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. This performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, an individuals entry into a high radiation area without proper knowledge of dose rates and radiological conditions placed the worker at risk for unnecessary radiation exposure. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance (Green)because:

(1) it was not an as low as is reasonably achievable (ALARA) finding,

(2) there was no overexposure,

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

(4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes [H12].

Enforcement.

Technical Specification 6.12.1 states, in part, that each high radiation area in which the intensity of radiation is greater than 100 millirem per hour, but less than 1,000 millirem per hour shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by issuance of a radiation work permit. Technical Specification 6.12.1(b) states, in part, that any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

Contrary to the above, on April 14, 2014, an individual entered a high radiation area, but was not on a radiation work permit that authorized entry into high radiation areas and was not knowledgeable of the dose rates in the area. Specifically, a security officer entered the shutdown heat exchanger room B, an established posted high radiation area, and his electronic dosimeter alarmed, measuring a peak dose rate of 107 millirem per hour. The Radiation Protection staff counseled the individual, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF-2014-01638. This violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report CR-WF-2014-01638, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy:

NCV 05000382/2014003-01; Failure to Control Entry into a High Radiation Area.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

The inspectors evaluated whether the licensee controlled in-plant airborne radioactivity concentrations consistent with ALARA principles and that the use of respiratory protection devices did not pose an undue risk to the wearer. During the inspection, the inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:

• The licensees use, when applicable, of ventilation systems as part of its engineering controls

• The licensees respiratory protection program for use, storage, maintenance, and quality assurance of National Institute for Occupational Safety and Health certified equipment, qualification and training of personnel, and user performance

• The licensees capability for refilling and transporting self-contained breathing apparatus air bottles to and from the control room and operations support center during emergency conditions, status of self-contained breathing apparatus staged and ready for use in the plant and associated surveillance records, and personnel qualification and training

• Audits, self-assessments, and corrective action documents related to in-plant airborne radioactivity control and mitigation since the last inspection These activities constitute completion of one sample of in-plant airborne radioactivity control and mitigation as defined in Inspection Procedure 71124.03.

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 Safety System Functional Failures (MS05)

a. Inspection Scope

For the period of January 1, 2013, through April 1, 2014, the inspectors reviewed licensee event reports, maintenance rule evaluations, and other records that could indicate whether safety system functional failures had occurred. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, and NUREG-1022, Event Reporting Guidelines: 10 CFR 50.72 and 50.73, Revision 3, to determine the accuracy of the data reported.

These activities constituted verification of the safety system functional failures performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Mitigating Systems Performance Index: Emergency AC Power Systems (MS06)

a. Inspection Scope

The inspectors reviewed the licensees mitigating system performance index data for the period of January 1, 2013, through April 1, 2014, to verify the accuracy and completeness of the reported data. 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 reported data.

These activities constituted verification of the mitigating system performance index for emergency ac power systems as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Mitigating Systems Performance Index: High Pressure Injection Systems (MS07)

a. Inspection Scope

The inspectors reviewed the licensees mitigating system performance index data for the period of January 1, 2013, through April 1, 2014, to verify the accuracy and completeness of the reported data. 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 reported data.

These activities constituted verification of the mitigating system performance indicator for high pressure injection systems as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.4 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors verified that there were no unplanned exposures or losses of radiological control over locked high radiation areas and very high radiation areas during the period of September 16, 2013, to March 31, 2014. The inspectors reviewed a sample of radiologically controlled area exit transactions showing no exposures greater than 100 millirem. 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 reported data.

These activities constituted verification of the occupational exposure control effectiveness performance indicator for occupational exposure control effectiveness as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.5 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual

Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed corrective action program records for liquid or gaseous effluent releases that occurred between September 16, 2013, and April 21, 2014, and were reported to the NRC to verify the performance indicator data. 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 reported data.

These activities constituted verification of the radiological effluent technical specifications offsite dose calculation manual radiological effluent occurrences performance index for high pressure injection systems 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 Semiannual Trend Review

a. Inspection Scope

The inspectors reviewed the licensees corrective action program, performance indicators, system health reports, operational plant issues lists, daily equipment status, and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors verified that the licensee was taking corrective actions to address identified adverse trends.

These activities constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

.3 Annual Follow-up of Selected Issues

a. Inspection Scope

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

• On December 4, 2013, public address system issue, malfunction during an evaluated exercise 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.

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

b. Findings

Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR Part 50.54(q)(2) because the licensee did not maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b).

Specifically, the licensee did not maintain the public address system in a manner that could provide protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10).

Description.

During the December 4, 2013, biennial exercise, NRC inspectors observed a failure of the public address system to provide audible announcements in the emergency facilities. This failure lasted for about one and one-half hours. Following the exercise, the licensee initiated an investigation into the cause and extent of condition of the failure. During discussion of investigation status between the licensee and the NRC inspectors on December 5, 2013, licensee staff mentioned that there were some public address system speakers that were non-functional prior to the exercise. The inspectors asked if the licensee had evaluated the cumulative effects of the non-functioning speakers (how many, how long they had been non-functioning, and if compensatory measures were put into place) on their ability to provide a range of protective actions for emergency workers and the public in accordance with 10 CFR Part 50.47(b)(10).

The licensee had not evaluated this in the past, so it was included in the scope of their ongoing investigation.

The investigation revealed the following:

• During the December 4, 2013, system failure, the public address system was inaudible in the service building, administration building, central alarm station, west side access, main support building, Emergency Operations Facility, and parts of the reactor auxiliary building and the turbine generator building.

• During the December 4, 2013, system failure, the public address system volume was lowered significantly in areas in the reactor auxiliary building, near the control room envelope; the Technical Support Center; and the Operations Support Center.

• Prior to the exercise, there were areas where the public address system speakers have been non-functional for a minimum of one year and six months. The total speakers affected are 30 out of 477 onsite (approximately 6 percent).

• Since the public address systems installation prior to 1984, the additional loads added to the system have resulted in system power demand during use that has exceeded the installed power amplifiers design capacity. During current public address system use, the demand is 1000 Watts per power amplifier.

The power amplifiers were only designed to supply 800 Watts (Condition Report CR-WF3-2013-05860).

• Changes in the public address system have been made that were not documented in the system engineering drawings. Additional speakers in newer buildings have been added that have not been documented (Condition Reports CR-WF3-2013-05879 and CR-WF3-2013-01823).

Condition reports associated with the public address system were reviewed over the current two-year period (from August 2011 through December 4, 2013). During that time, there were numerous component failures providing evidence that the systems adequacy was not being maintained. Examples include:

• Eight instances involving the receipt of the systems supervisory alarm in the control room, indicating electrical component failures in the system (Condition Reports CR-WF3-2012-02055, CR-WF3-2012-03319, CR-WF3-2012-05959, CR-WF3-2012-07316, CR-WF3-2013-00752, CR-WF3-2013-05506, CR-WF3-2013-05594, and CR-WF3-2013-05751)

• Three failures to function during Emergency Preparedness drills (Condition Reports CR-WF3-2012-03566, CR-WF3-2013-01824, and CR-WF3-2013-05594)

• Two examples of failures during actual site medical events that delayed emergency response (Condition Reports CR-WF3-2012-05959 and CR-WF3-2013-03840)

The licensee developed an engineering proposal in August 2002 to address noted deficiencies in the public address system. In Engineering Request ER-W3-1998-0208-000, dated August 21, 2002, the licensee indicated that the state of the system involved paging trains and the amplifiers operating in overloaded conditions, including the portion of the public address system in the turbine generator building being extremely overloaded. The proposal denotes that the systems amplifier manufacturer did not make the components anymore, there was need to add additional system loads to provide paging to other areas, and the power supply for the system (PDP 3014AB) was fully loaded. In the Design Input Record attachment, it states, The paging amplifiers have been experiencing failure due to load exceeding the capacity of the installed. In addition, the equipment is most probably approaching end of operating life with no replacement models as the original manufacturer is no longer in business.

The licensees plan was to upgrade the public address system, its power supply, and add additional needed loads. Engineering Request ER-W3-2000-0086-000-00 documented the plan to install permanent public address system loads in the construction support building during the same timeframe. Based on the licensees review of these engineering requests documented in Condition Report CR-W3-2013-05860, and confirmed via discussion with licensee staff on February 26, 2013, the plans to modify the system to address its reliability issues was never completed. The licensee added additional permanent loads to the system in the construction support building without addressing the overload and reliability concerns previously identified.

Condition reports reviewed did not include an evaluation of the public address systems functional capability versus all applicable emergency planning requirements.

For example, Condition Reports CR-WF3-2012-02055, CR-WF3-2012-05959, and CR-WF3-2013-00752 stated common language in the Operability section, saying that the system component failures did not require an operability determination or a functional assessment. This was decided because the system was categorized as non-safety-related, so the evaluation was not applicable. While the public address system is nonsafety-related, it does provide the means to make a range of protective actions available to the workers and the public in accordance with 10 CFR Part 50.47(b)(10).

The specific details on how the licensees public address system meets this function are documented in the licensees emergency plan. With no evidence that this functional assessment had been completed at any time, on February 3, 2014, the inspectors asked the licensee how the public address system, with its repeated component failures, was in compliance with the site emergency plan.

The current state of the licensees public address system is not in compliance with their emergency plan. In the Waterford 3 SES Emergency Plan, Revision 044, Section 7.5.1.2, it states, in part, that a malfunction of a power amplifier will only affect 25 percent of the speakers, but in no case will an area or building lose total coverage.

Also, in Section 7.5.3.1, it discusses the function of the fire and station alarms, which transmit their signal via the public address system to the site. These alarms, credited as making up the onsite emergency notification system, are conveyed through paging amplifiers and broadcast through the loudspeaker system covering the entire site.

During the December 4, 2013, exercise, the public address system did not function in multiple entire buildings. In addition, with the record of public address system failures documented in condition reports and the August 2002 engineering request document, this resulted in multiple instances when the credited fire and station alarms would not have been heard throughout the entire plant site had an actual emergency occurred.

Resulting from the functionality issues with the public address system, the licensee is implementing compensatory measures until the system is restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications will be sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation.

In summary, during the period from August 2002 to present, the licensee made changes to the public address system that have degraded their ability to use the system to implement a range of protective actions to protect emergency workers onsite, as required by 10 CFR Part 50.47(b)(10). This resulted in failure to maintain emergency plan compliance, which culminated in the failure of the public address system for approximately one and a half hours during the December 4, 2013, biennial emergency preparedness evaluated exercise.

Analysis.

The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. The inspectors performed the significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings (issue date: June 19, 2012); and Appendix B, Emergency Preparedness Significance Determination Process (SDP), Attachment 2 (issue date: February 24, 2012), to evaluate the issue. Because this issue dealt with both a percentage of public address speakers not functioning for an extended time period, as well as the long-term degraded condition of public address system components, no significance example fully enveloped the finding being considered within Table 5.10-1 of Appendix B. Therefore, per Section 5.0.1(c) of Appendix B, the finding was compared against the definitions with Appendix B, Attachment 2. The inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function.

The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.

From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes [P.2].

Enforcement.

Title 10 CFR Part 50.54 (q)(2) requires, in part, a licensee to follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part and, for nuclear power reactor licensees, the planning standards of 10 CFR 50.47(b). Part 50.47(b)(10) requires licensees to develop a range of protective actions for the plume exposure pathway emergency planning zone (EPZ) for emergency workers and the public. Waterford 3 SES Emergency Plan, Revision 044, establishes the onsite loudspeaker system as the means to satisfy this requirement.

Section 7.5.3.1 of the Plan discusses the function of the fire and station alarms, which transmit their signal via the public address system to the site. These alarms, credited as making up the onsite emergency notification system, are conveyed through paging amplifiers and broadcast through the loudspeaker system covering the entire site.

In addition, Section 7.5.1.2, places a limit on how extensive loudspeaker malfunctions can be and states, in part, that a malfunction of a power amplifier will only affect 25 percent of the speakers, but in no case will an area or building lose total coverage.

Contrary to the above, from August 21, 2002, to December 4, 2013, the licensee failed to maintain the effectiveness of the emergency plan, in particular, with reference to its compliance with 10 CFR Part 50.47(b)(10). Specifically, the licensee failed to maintain the public address system in a manner to ensure no areas or buildings would lose total coverage, nor that the credited fire and station alarms would be conveyed via the public address system to the entire site. The licensee has compensatory measures in place, and entered the issue into the corrective action program in Condition Report CR-W3-2013-05860.

Because the licensee entered the issue into its corrective action program and the finding is of very low safety significance (Green), this violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the NRCs Enforcement Policy:

NCV 05000382/2014003-02, Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions.

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

(Closed) Licensee Event Report 05000382/2014-002-00, Inadequate Tightening of Starting Air Filter Housing results in Inoperable Train A Emergency Diesel Generator On March 1, 2014, the licensee identified that the filter housing cover on the emergency diesel generator train A starting air filter had unfastened from its base. The emergency diesel generator is started by means of compressed air. The air filter is in the flowpath to deliver compressed air to one bank of the emergency diesel generator. With an air leak at the filter due to the filter housing being loose or not installed, air would likely not be delivered to the corresponding air distributor, and thus one bank of the emergency diesel generator would not receive compressed air to start the emergency diesel. The last successful start of the emergency diesel generator A was on February 27, 2014. It was assumed that the cover became unfastened during that run; therefore, the emergency diesel generator was potentially inoperable for 2 days as a result of this condition. The licensee identified that the actions required by technical specifications for this condition had not been performed. Specifically, test of emergency diesel generator B was not performed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and the requirement to demonstrate the operability of the remaining alternating current circuits at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or to be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> was not performed.

Upon discovery of this condition, the licensee appropriately entered the applicable technical specifications and followed the associated actions. The filter housing was promptly replaced by the operator who discovered the condition, restoring emergency diesel generator A to an operable status. In addition, the licensee verified that the other three filters on emergency diesel generator A and B starting air systems were properly fastened. The inspectors and the licensee reviewed the associated technical manuals and determined that there were not any specified torque values for this component. In the review of this event, no findings or violations of NRC requirements were identified.

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

4OA5 Other Activities

.1 (Closed) URI 05000382/2013005-02 Protective Action Recommendations Under

Conditions of Changing Wind Vectors Not Consistent with Federal Guidance The inspectors identified, during the December 4, 2013, evaluated exercise, an unresolved item related to the adequacy of the licensees guidelines for the choice of protective actions during an emergency in accordance with the requirements of 10 CFR 50.47(b)(10). The issue was identified as an unresolved item because the NRC had not determined whether the licensee had adequately implemented Planning Standard 10 CFR 50.47(b)(10), which states, in part, that guidelines for the choice of protective actions, consistent with Federal Guidance, are developed and in place.

Specifically, the NRC had not determined whether the restrictions on the application of radiological assessments in Procedure EP-002-052, "Protective Action Guidelines,"

Revision 23, Step 5.4.1.1, adequately implement the guidance of Environmental Protection Agency (EPA) document, EPA-400-R-92-001, "Manual of Protective Action Guides and Protective Action for Nuclear Incidents." On June 23, 2014, the licensee provided information indicating that, at the time of the exercised protective action recommendation (PAR), General Emergency conditions continued to exist. The NRC determined that the PAR was appropriate to the circumstances presented in the exercise and that no performance deficiency existed.

.2 Follow Up Inspection for Three or More Severity Level (SL) IV Traditional Enforcement

Violations in the Same Area in a 12-Month Period (IP 92723)

a. Inspection Scope

The NRC performed Inspection Procedure 92723, Follow Up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period. The licensee received three Severity Level IV violations in the traditional enforcement area of impeding regulatory process at the end of the 2013 mid-cycle assessment period. The licensee informed the NRC on November 13, 2013, of their readiness to perform this follow-up activity. The NRC reviewed the licensees corrective action documents for each violation and the overall cause analysis for the following items:

• Problem identification

• Cause, extent of condition and extent of cause

• Evaluation of corrective actions

b. Findings and Observations

No findings were identified.

The inspectors determined that the licensee properly identified the problem and causes using a systematic approach. The evaluation adequately addressed the extent of condition and extent of cause. It appears that the corrective actions taken or planned were appropriate to address the causes, and schedules to measure the success of these actions were established.

.3 Follow-up on Traditional Enforcement Actions Including Violations, Deviations,

Confirmatory Action Letters, Confirmatory Orders, and Alternative Dispute Resolution Confirmatory Orders (IP 92702)a. Background On August 24, 2011, the NRC issued a Confirmatory Order (EA-11-096) to Entergy Operations, Inc., and Entergy Nuclear Operations, Inc. (collectively referred to as Entergy). The Confirmatory Order actions were agreed upon by Entergy and the NRC during an alternative dispute resolution session held on July 18, 2011, to resolve NRC concerns regarding an apparent violation of employee protection requirements at the River Bend Station. The actions focused on reorganizing the Quality Control reporting relationships, ensuring adequate training of 10 CFR 50.7, Employee Protection, and performing an effectiveness review of the Employee Concerns Program procedures at all Entergy facilities.

By letter dated August 23, 2012, Entergy notified the NRC of the actions that had been taken in response to the requirements imposed by the Confirmatory Order. Accordingly, during the week of April 29, 2013, NRC staff from the Office of Enforcement and Region IV performed an inspection at the River Bend Station to assess the specific actions identified in Entergys response letter. NRC staff also verified implementation of the remaining actions required to satisfy the conditions set forth in the Confirmatory Order for all Entergy sites. Subsequent to this inspection, NRC staff continued to interact with Entergy regarding the adequacy of the corrective and preventive actions related to the underlying discriminatory issue.

b. Findings and Observations

No findings were identified.

During the follow-up inspection, the NRC staff reviewed Entergys Employee Concerns Program supervisory training and general employee training documents, the relevant lessons learned from the facts of this matter, and the fleet-wide written communication reinforcing Entergys commitment to maintaining a safety-conscious work environment.

The NRC staff also reviewed the General Employee Training and Supervisory Training modules. Based on these reviews, it was determined that these training modules adequately addressed employee protection and included insights from the underlying discriminatory matter. The NRC staff determined that the supervisory training module appeared complete and included case studies as well as the specific elements from the underlying § 50.7, Employee Protection, violation. However, it was noted that although employees receive General Employee Training on an annual basis, Entergy does not require supervisors to take employee protection refresher training on a recurring basis, as a means to reinforce these standards.

Additionally, NRC staff evaluated the results of Entergys effectiveness review of Employee Concerns Program (ECP) enhancements and the associated training that arose from the corrective actions taken to address this matter. Based on the results of this evaluation, it was determined that Entergy had performed the requisite reviews at each station including examination of selected ECP Case Files, Records Retention, Concerned Individual follow-up, and ECP Coordinator training. Within the areas examined, no findings were identified and in general it was determined that Entergy had adequately performed the effectiveness review of ECP procedural enhancements and the ECP training related to this matter.

During the follow-up review of the Quality Control/Quality Assurance reporting relationship, it was determined that Entergys response did not ensure that persons performing the quality assurance function of receipt inspection reported to a management level sufficient to maintain organizational freedom and independence from cost and schedule were maintained. Subsequent to the identification of this performance issue, which affected the implementation of the quality assurance program at all nine Entergy sites, the condition was entered into the licensees corrective action program as Condition Report CR-HQN-2013-00466.

Following the identification of this issue, additional discussions were held between NRC and Entergy to clarify the intent of the settlement agreement and subsequent Confirmatory Order stemming from the earlier alternate dispute resolution mediation.

As a result of these discussions, Entergys Corporate Licensing organization developed a fleet reconciliation plan to modify Entergys Quality Assurance Program Manual to require that individuals performing inspections in accordance with Quality Assurance Program Manual, Section B.12, Inspection, functionally report to the associated manager responsible for Quality Assurance. As described in the corrective actions associated with Condition Report CR-HQN-2013-00466, the affected individuals were those requiring certification in accordance with Quality Assurance Program Manual, Table 1, Regulatory Commitments, Section G, Regulatory Guide 1.58, Revision 1, Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel, dated September 1980. In addition to revising the applicable provisions in the Quality Assurance Program Manual, corrective actions were initiated to revise implementing procedures to reflect the change in reporting relationship during the performance of required inspections as well as providing training to the affected individuals. The NRC staff confirmed that the remaining conditions of the Confirmatory Order were adequately addressed.

Based on the above reviews, the NRC determined that Entergy properly implemented the conditions specified in the Confirmatory Order and that the associated actions were adequately implemented.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On April 1, 2014, the inspectors presented the problem identification and resolution (annual follow-up of selected issues) inspection results to Mr. C. Rich, Jr., Director, Regulatory and Performance Improvement, 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 24, 2014, the inspectors presented the radiation safety inspection results to Mr. M. Chisum, 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.

On May 1, 2014, the inspectors presented the inservice inspection results to Mr. M. Chisum, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors returned copies of proprietary documents to the licensee.

On June 24, 2014, the inspector presented the results of the in-office inspection of Unresolved Item 05000382/2013005-02 to Mr. C. Rich, Jr., Director, Regulatory and Performance Improvement, and other members of the licensee staff. The licensee acknowledged the issues presented.

On July 1, 2014, the inspectors presented the integrated inspection results to Mr. M. Chisum, 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

M. Chisum, Site Vice President

K. Cook, General Manager, Plant Operations

E. Brauner, Manager, Emergency Preparedness

J. Briggs, Superintendent, Electrical Maintenance

J. Cary, Supervisor, Radiation Protection

M. Chaisson, Supervisor, Radiation Protection

K. Crissman, Senior Manager, Maintenance

D. Frey, Manager, Radiation Protection

B. Ford, Senior Manager Nuclear Safety and Regulatory Assurance

R. Gilmore, Manager, Systems and Components

M. Haydell, Engineering Supervisor

A. James, Manager, Security

J. Jarrell, Manager, Regulatory Assurance

B. Lanka, Director, Engineering

N. Lawless Manager, Chemistry

B. Lindsey, Senior Manager, Operations

W. McKinney, Manager, Performance Improvement

S.W. Meiklejohn, Superintendent, I & C Maintenance

M. Mills, Manager, Nuclear Oversight

L. Milster, Licensing Specialist, Regulatory Assurance

R. OQuinn, Senior Staff Engineer

B. Pellegrin, Senior Manager, Production

G. Pickering, Senior Engineer

J. Pollock, Licensing Specialist, Regulatory Assurance

R. Porter, Manager, Design & Program Engineering

D. Reider, Supervisor, Quality Assurance

C. Rich, Jr., Director, Regulatory and Performance Improvement

R. Sherman, Supervisor, Radiation Protection

J. Signorelli, Acting Manager, Training

R. Simpson, Superintendent, Operator Training

P. Stanton, Supervisor, Design Engineering

J. White, Supervisor, Radiation Protection

J. Williams, Senior Licensing Specialist

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000382-2014003-01 NCV Failure to Control Entry into a High Radiation Area (Section 2RS1)

05000382-2014003-02 NCV Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions (Section 4OA2.2)

Closed

05000382-2014-002-00 LER Inadequate Tightening of Starting Air Filter Housing results in Inoperable Train A Emergency Diesel Generator (Section 4OA3)

05000382/2013005-02 URI Protective Action Recommendations Under Conditions of Changing Wind Vectors Not Consistent with Federal Guidance (Section 4OA5)