IR 05000382/2014005

IR 05000382/2014005; on 10/01/2014 - 12/31/2014; Waterford, Unit 3, Adverse Weather Protection, Operability Determinations and Functionality Assessments, Post-Maintenance Testing, Occupational ALARA Planning and Controls, and Other Activiti

ML15044A273
Person / Time
Site:	Waterford
Issue date:	02/12/2015
From:	Ryan Lantz NRC/RGN-IV/DRP/RPB-E
To:	Chisum M Entergy Operations
Lantz R
References
IR 2014005
Download: ML15044A273 (56)

Text

UNITED STATES ary 12, 2015

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2014-005

Dear Mr. Chisum:

On December 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Waterford Steam Electric Station, Unit 3. On January 15, 2015, 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 six findings of very low safety significance (Green) in this report.

Five of these findings involved violations of NRC requirements. Further, the NRC inspectors documented a licensee-identified violation which was determined to be Severity Level IV significance in this report. 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.

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. 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/

Ryan Lantz, Chief Project Branch E Division of Reactor Projects Docket Nos. 50-382 License Nos. NPF-38

Enclosure:

Inspection Report 05000382/2014-005 w/ Attachment: Supplemental Information

REGION IV==

Docket: 05000382 License: NPF-38 Report: 05000382/2014005 Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: 17265 River Road Killona, LA 70057 Dates: October 1 through December 31, 2014 Inspectors: F. Ramírez, Senior Resident Inspector C. Speer, Resident Inspector P. Elkmann, Senior Emergency Preparedness Inspector N. Greene, PhD, Health Physicist P. Hernandez, Health Physicist Approved Ryan Lantz By: Chief, Project Branch E Division of Reactor Projects-1- Enclosure

SUMMARY

IR 05000382/2014005; 10/01/2014 - 12/31/2014; Waterford Steam Electric Station, Unit 3;

Adverse Weather Protection, Operability Determinations and Functionality Assessments,

Post-Maintenance Testing, Occupational ALARA Planning and Controls, and Other Activities.

The inspection activities described in this report were performed between October 1 and December 31, 2014, by the resident inspectors at the site and inspectors from the NRCs Region IV office. Six findings of very low safety significance (Green) are documented in this report. Five of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented one licensee-identified violation of Severity Level IV significance in this report. 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 Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 312, in two separate instances.

Specifically, on both November 16 and December 23, 2014, the licensee entered the off-normal procedure due to a tornado watch, but failed to assess and control potential tornado-borne missile hazards on site as required by the procedure. The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-05912 and CR-WF3-2014-06453. The immediate corrective action taken to restore compliance was to secure the identified hazards.

This finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in the event of a tornado at the site, these loose items could have become missiles with the potential to impact safety-related site equipment and personnel. The inspectors determined the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or functions specifically designed to mitigate a seismic, flooding, or severe weather event (e.g. seismic snubbers, flooding barriers, tornado doors). The inspectors concluded that the finding had a cross-cutting aspect in the area of Human Performance, Field Presence, because the licensee did not ensure supervisory and management oversight of work activities [H.2] (Section 1R01).

Green.

The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to assess immediate operability of safety-related systems in accordance with site procedures, in three separate instances. Specifically, on two occasions, the licensee did not properly assess operability of safety-related relays in the Engineered Safety Features Actuation Signal system, which in turn brought into question the operability of the emergency diesel generators. A third example involved the licensees failure to properly assess operability of safety-related class 3 piping in the dry cooling towers, which brought into question the operability of the component cooling water system. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-06014. The licensee restored compliance by revising the immediate operability determinations to reflect an adequate reason to justify operability of the systems in question.

The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to follow the Operability Determination procedure caused the licensee to incorrectly assess the capability of the systems impacted by the relays and dry cooling tower tube leak to perform their safety function and there was a reasonable doubt on the operability of the systems. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process, because individuals did not use a consistent, systematic approach to make a decision, and risk insights were not incorporated appropriately [H.13] (Section 1R15).

Green.

The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for failure of the licensee to develop a preventative maintenance schedule for inspections of safety-related equipment. Specifically, the licensee did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04930 and CR-WF3-2014-06100. The licensee developed preventative maintenance tasks to inspect the dry cooling tower tubes, including disassembly where necessary, to restore compliance.

The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to inspect portions of the dry cooling towers prevented the licensee from identifying corrosion that eventually degraded the system enough to cause a leak. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. The inspectors concluded that the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee did not implement a corrective action program with a low threshold for identifying issues [P.1]

(Section 1R19).

Green.

The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, for the licensees failure to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not have an adequate replacement frequency for safety-related relays associated with engineered safety features equipment to ensure that all required equipment operated in the time sequence assumed by the safety analysis. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-05091. The licensee replaced the affected relays and reduced their replacement frequency from 18 years to 3 years to restore compliance.

The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop an adequate replacement frequency for the relays used to monitor for under-voltage conditions on the safety-related emergency busses could have prevented the equipment from performing its safety function. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system component and the affected equipment maintained its operability. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, Challenging the Unknown, because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding [H.11] (Section 1R19).

Green.

The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, Corrective Actions, for the licensees failure to correct a condition adverse to quality in a time commensurate with the safety significance of the issue. Specifically, the licensee failed to repair degraded conduit that had been identified as corroded since 2008.

As a result, conduits that were housing cables for safety-related components were degraded to the point where water entered the conduit and submerged cables that were not designed for submergence for an extended period of time. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04951. The licensee repaired the degraded conduit associated with the impacted safety-related cables to restore compliance, and also initiated an extent of condition review to identify other cables that could potentially be impacted by degraded conduits.

The inspectors determined that the performance deficiency was more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, safety-related cables that were not rated for full submergence were submerged in water since at least 2008, potentially affecting the integrity of the cable and potentially impacting the safety-related equipments ability to perform their safety function in the event of an accident. The inspectors determined that the finding had very low safety significance (Green) because the finding impacted the qualification of mitigating components, but the components maintained operability. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensees decision-making practices did not emphasize prudent choices over those that are simply allowable. Specifically, when evaluating condition reports written through several years that document the degraded conduit, the licensee elected to defer needed maintenance instead of placing the adequate priority on the issue. [H.14]

(Section 4OA2).

Cornerstone: Occupational Radiation Safety

Green.

The inspectors identified a finding associated with the licensees failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations during Refueling Outage 19. Specifically, the inspectors concluded that, had the licensee appropriately evaluated the Alloy 600 pipe weld conditions/locations during the ALARA planning process and appropriately performed in-progress ALARA reviews, they could have reasonably planned for the full scope of work and provided a better estimate and/or adequately justified revising the estimate for the job. These failures to plan and control the job activities led to unplanned, unintended collective dose. The licensee evaluated the procedures used during this work, including their process for planning and estimating doses, and documented the issue in the corrective action program.

The failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations is a performance deficiency. This performance deficiency is more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, it caused the collective radiation dose for the work to be greater than 5 man-rem and exceed the planned dose estimate by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding has very low safety significance because: (1) it was associated with ALARA planning and (2) the licensees three-year rolling average collective dose of 121.7 man-rem was less than 135 man-rem. The finding has a Work Management cross-cutting aspect, associated with the Human Performance cross-cutting area, because the licensee did not adequately plan or control work activities such that nuclear safety is the overriding safety priority.

Specifically, the ALARA plan did not reflect the time needed to complete the work activities, thus underestimating the dose requirements, and the administrative control of reviewing the work-in-progress at appropriate completion points failed. [H.5] (Section 2RS2)

Licensee-Identified Violations

A Severity Level IV violation that was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

PLANT STATUS

The Waterford Seam Electric Station, Unit 3, began the inspection period at 100% power and maintained 100% power for the duration of the period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On November 16, 2014, and on December 23, 2014, the National Weather Service declared a tornado watch in the vicinity of the facility. 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 the licensees 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 constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

b. Findings

Failure to Identify and Control Potential Tornado-Borne Missile Hazards

Introduction.

The inspectors identified a Green, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 312, in two separate instances. Specifically, on both November 16 and December 23, 2014, the licensee entered the off-normal procedure due to a tornado watch, but failed to assess and control potential tornado-borne missile hazards on site as required by the procedure.

Description.

On November 16 and again on December 23, 2014, a tornado watch was issued for the area surrounding the Waterford Unit 3 site. On both occasions, operations personnel directed maintenance personnel to tour the area surrounding the site and to secure or store any loose items that could become tornado-generated missile hazards in accordance with off-normal procedure OP-901-521, Severe Weather and Flooding, Revision 312.

The inspectors reviewed the requirements of OP-901-521 and procedure EN-FAP-EP-010, Severe Weather Response, which is referenced by OP-901-521 for examples of potential tornado-generated missiles to secure. The inspectors toured the plant site during each tornado watch to verify the licensees implementation of OP-901-521 to ensure all tornado doors were functional and potential tornado-borne missile hazards were appropriately identified and controlled. On each plant tour, the inspectors identified numerous areas with loose material which could become potential tornado-borne missile hazards, several of which corresponded to examples given in EN-FAP-EP-010 of loose items to secure during a tornado watch. Examples of loose materials included unsecured scaffolding, uncovered garbage in large dumpsters, unsecured trash bins, and other loose metallic items. In the event of a tornado, the loose items could have become missiles with the potential to impact safety-related site equipment and personnel.

The inspectors also reviewed the completion of OP-901-521, Attachment 4, Severe Weather Equipment Actions and Restoration, which requires the licensee to log equipment or loose items that need to be placed in a safe condition in the event of severe weather. During the November 16, 2014, walkdown, the inspectors noted that the attachment had been signed off as completed by operations personnel by the time the inspectors completed the inspection. As a result, the items identified by the inspectors would have been left uncorrected. Following the site walkdowns, the inspectors notified operations personnel of the concerns and the licensee took action to address them.

Analysis.

The inspectors concluded that the failure to assess and control potential tornado-borne missile hazards as required by licensee procedure OP-901-521, Severe Weather and Flooding, Revision 312 was a performance deficiency. The inspectors determined that the performance deficiency was reasonably within the licensees ability to foresee and correct. The inspectors concluded that the performance deficiency was more than minor, and therefore a finding, because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, in the event of a tornado at the site, the loose items could have become missiles with the potential to impact safety-related site equipment and personnel.

The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate the finding for its impact on the Mitigating Systems Cornerstone. The initial screening directed the inspectors to use Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the finding. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not involve the loss or degradation of equipment or functions specifically designed to mitigate a seismic, flooding, or severe weather event (e.g. seismic snubbers, flooding barriers, tornado doors).

The inspectors concluded that the finding reflected current licensee performance and had a cross-cutting aspect in the area of Human Performance, Field Presence, because the licensee did not ensure supervisory and management oversight of work activities

[H.2].

Enforcement.

Technical Specification 6.8.1.a, requires, in part, that procedures shall be established, implemented and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6.w of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combatting Acts of Nature, including tornados. The licensee established procedure OP-901-521, Severe Weather and Flooding, Revision 312, to meet this requirement. OP-901-521 requires loose items that pose a threat to the plant equipment or personnel be secured during a tornado watch or warning.

Contrary to the above, on November 16, 2014, and on December 23, 2014, the licensee failed to implement procedure OP-901-521, Severe Weather and Flooding, Revision 312. Specifically, during a tornado watch, the licensee did not secure loose items that, in the event of a tornado, could have become missiles and posed a threat to the plant equipment or personnel. The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-05912 and CR-WF3-2014-06453. The immediate corrective action taken to restore compliance was to control the identified hazards.

Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation is treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000382/2014005-01, Failure to Identify and Control Potential Tornado-Borne Missile Hazards.

.2 Readiness to Cope with External Flooding

a. Inspection Scope

On December 4, 2014, the inspectors completed an inspection of the stations readiness to cope with external flooding. After reviewing the licensees flooding analysis, the inspectors chose the dry cooling tower area to inspect.

The inspectors reviewed plant design features and licensee procedures for coping with flooding. The inspectors walked down the dry cooling tower area to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether credited operator actions could be successfully accomplished.

These activities constituted one sample of readiness to cope with external flooding, 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 October 29, 2014, chemical and volume control system train B with train A out-of-service for maintenance

• On November 12, 2014, essential chilled water system train AB following system realignment

• On December 2, 2014, emergency diesel generator B with emergency diesel generator A out-of-service for unplanned maintenance 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 or trains were correctly aligned for the existing plant configuration.

These activities constituted three 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 four plant areas important to safety:

• On October 3, 2014, fire area RAB 2, main steam isolation valve passage

• On October 3, 2014, fire area RAB 3, heating, ventilation and air conditioning (HVAC) equipment room

• On December 2, 2014, fire area RAB 17, component cooling water heat exchanger B

• On December 2, 2014, fire area RAB 32, auxiliary component cooling water room and pipe penetration area 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 four quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

On November 10, 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 EPRI NP-7552 for the dry cooling towers and the material condition of the heat exchanger internals. Additionally, the inspectors walked down the dry cooling towers to observe its material condition and verified that the dry cooling towers were correctly categorized under the Maintenance Rule and were 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.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On December 3, 2014, the inspector observed simulator training for an operating crew.

The inspector assessed the performance of the operators and the evaluators critique of their performance.

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 December 16, 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 risk due to maintenance activities associated with a switchgear air handling unit. The inspectors observed the operators performance of the following activities:

• Alarm response

• Shift briefings 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 (SSCs):

• On October 22, 2014, containment purge radiation monitors

• On December 22, 2014, safety injection tanks 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 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

On November 17, 2014, the inspectors reviewed a risk assessment performed by the licensee prior to an auxiliary feedwater pump outage and the risk management actions taken by the licensee in response to elevated risk.

The inspectors verified that this risk assessment was 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 assessment and verified that the licensee implemented appropriate risk management actions based on the result of the assessment.

The inspectors also observed portions of two emergent work activities that had the potential to cause an initiating event and to affect the functional capability of mitigating systems.

• On November 2, 2014, emergent maintenance on auxiliary component cooling water header A component cooling water heat exchanger outlet temperature control valve

• On November 18, 2014, emergent maintenance on startup transformer A 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 (SSCs).

These activities constitute completion of three 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 three operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components (SSCs):

• On October 2, 2014, auxiliary component cooling water pump B

• On October 14, 2014, ultimate heat sink with an isolated dry cooling tower tube bundle

• On October 16, 2014, emergency diesel generators due to potential 4KV under-voltage relay failures 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 the degraded SSC.

These activities constitute completion of three operability and functionality review samples, as defined in Inspection Procedure 71111.15.

b. Findings

Failure to Follow the Operability Determination Process

Introduction.

The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to assess immediate operability of safety-related systems in accordance with site procedures, in three separate instances. Specifically, on two occasions, the licensee did not properly assess operability of safety-related relays in the engineered safety features actuation signal system, which in turn brought into question the operability of the emergency diesel generators. The third example involved the failure to properly assess operability of safety-related class 3 piping in the dry cooling towers, which brought into question the operability of the component cooling water system.

Description.

On October 2014, the licensee sent a report to the NRC in accordance with 10 CFR 21, Reporting of Defects and Noncompliance, due to the deviation of a dedicated basic component from manufacturing specifications. The Part 21 report was associated with Allen Bradley relays installed in engineered safety features actuation signal system applications. The licensees independent failure analysis determined that due to corrosion the relays could fail and in turn impact the component being supplied by the relay. When the licensee evaluated the extent of condition, they concluded that a total of 33 relays could be impacted by this failure mechanism. Since most of these 33 Allen Bradley relays would fail in their safety position, the inspectors operability review focused on four relays that monitored the safety-related 4kV bus for an under-voltage condition.

The licensees evaluation for these four relays concluded that the failure mode could cause the under-voltage monitoring relays to initiate a diesel sequencer lock-out if the failure mode occurred concurrent with a loss of offsite power or concurrent with a safety injection actuation signal. This meant that the emergency diesel generator sequencer would not be able to complete its sequencing function until it was reset. The emergency diesel generators were susceptible to this failure in the 192-second window where the sequencing of loads was occurring.

The licensee evaluated the operability of the impacted systems as part of the Operability Determination Process. When the inspectors reviewed the licensees operability determination, they noted that the engineering evaluation which supported the operability determination stated that, based on the failure mode being unlikely to occur within the 192-second time period of the emergency diesel generator sequencer, there was reasonable assurance that the systems supplied by these Allen Bradley relays (i.e. the diesel generators) remained capable of performing their safety function. Operations personnel, in turn, accepted this evaluation that was based on a probability statement and considered the emergency diesel generators operable. Additionally, as part of the established corrective actions, the engineering department was provided additional time to further evaluate the issue, determine failure rates, and evaluate whether any compensatory measures were needed.

The inspectors noted that licensee procedure EN-OP-104, Operability Determination Process, Revision 7, stated that it is not acceptable to use probabilistic risk assessment for making operability determinations. In addition, the inspectors consulted NRC Inspection Manual Chapter 0326, Operability Determinations and Functionality Assessments for Conditions Adverse to Quality or Safety, which is considered the NRC staffs position on assessing system operability. The manual chapter states that the use of PRA or probabilities of occurrence of accidents or external events is not consistent with the assumption that the event occurs, and is not acceptable for making operability decisions.

Since the relay failure mode could occur and prevent the emergency diesel generators from performing their safety function, basing the operability on the likelihood of a failure was not in accordance with established guidelines. When the inspectors brought this issue to the licensees attention, since the operability of the diesels was called into question, the licensee reassessed operability and documented a different acceptable reason to justify operability of the emergency diesel generators.

In October 2013, the licensee had identified a different failure mode for the same set of Allen Bradley relays. At the time, it had been reported to the NRC on a separate 10 CFR 21 report that these Allen Bradley relays could fail due to spurious de-energization. Even though it was a different failure mechanism, the impact on the emergency diesel generators would have been the same as described earlier. The inspectors reviewed the operability determination written in October 2013. The inspectors noted, as a second example, that the licensee had used the same statement to justify operability of the emergency diesel generators. Specifically, the operability determination stated that because the failure mechanism was unlikely to occur, the emergency diesel generators were able to perform their safety function. However, that operability determination was later revised to reflect a different and acceptable justification for operability and it did not require inspector intervention.

The third example where the licensee did not properly assess the operability of a system occurred in February 2014. The licensee identified a leak in the dry cooling tower tube bundles, which provide cooling for the component cooling water system.

The dry cooling tower tube bundles are considered ASME Class 3 piping. As required by procedure EN-OP-104, and consistent with NRC guidance, the component must be evaluated in an immediate determination of operability to support a reasonable expectation of operability. Procedure EN-OP-104 further stated that the degradation mechanism would have to be readily apparent to support a determination of operable and that to be readily apparent the degradation must be discernible from visual inspection.

However, the licensee declared the component cooling water system operable at the time based on evaluating the different potential failure modes and not by specifically identifying the cause of the leak. In addition, a visual inspection was not performed. Since the dry cooling tower tube bundles were repaired before the inspectors identified this example, current operability of the system was not a concern.

Analysis.

The inspectors determined that the failure to perform an adequate immediate operability determination of the safety-related 4KV under-voltage monitoring relays and the dry cooling tower tube bundles in accordance with procedure EN-OP-104, Operability Determination Process, was a performance deficiency that warranted further evaluation.

Using the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, failing to follow the Operability Determination Process caused the licensee to incorrectly assess the capability of the impacted safety-related systems to perform their safety function, which resulted in a reasonable doubt on the operability of the systems.

Further, it required inspector intervention for the licensee to re-assess and revise the operability determination associated with the Allen Bradley relays.

Using Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the finding was determined to have very low safety significance (Green)because it did not affect the design or qualification of the systems, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment.

This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process, because individuals did not use a consistent, systematic approach to make a decision, and risk insights were not incorporated appropriately. Specifically, the Operability Determination Process was not being systematically applied, and, as a result, systems could potentially be incorrectly evaluated as operable [H.13].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and be accomplished in accordance with these instructions, procedures, or drawings. Licensee procedure EN-OP-104, Operability Determination Process, Revision 7, provides a process to assess operability and functionality when degraded or nonconforming conditions affecting structures, systems, and components are identified.

Procedure EN-OP-104, states, in part, that it is not acceptable to use probabilistic risk assessment for making operability determinations. In addition, procedure EN-OP-104 states, in part, that upon discovery of leakage from an ASME Class 2 or Class 3 safety system or component, the component must be evaluated in an immediate determination of operability to support a reasonable expectation of operability. Procedure EN-OP-104 further states that, in performing the immediate operability determination, the degradation mechanism would have to be readily apparent to support a determination of operable.

Contrary to the above, in October 2013, and October 2014, the licensee used probabilistic risk assessment for making operability determinations. Specifically, when a potential failure mode was discovered for the safety-related relays that monitor the 4kV bus for an under-voltage condition, and the operability of the emergency diesel generators was called into question, the licensee stated that since the failure was unlikely to occur, there was reasonable assurance that the system could perform its safety function. In addition, in February 2014, the licensee discovered a leak in the dry cooling tower tube bundles, which are classified as Class 3 piping, but did not determine the degradation mechanism to support the operability determination by which the system was declared operable. After NRC inspector questioning, the licensee re-assessed operability of these safety-related systems appropriately.

The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-06014. The immediate corrective action taken to restore compliance for the evaluation associated with the 4KV under-voltage monitoring relays included revising the immediate operability determination to reflect an adequate reason to justify operability of the systems in questions. The dry cooling tower tube bundles were already repaired when the inspectors identified the improper operability determination. Corrective actions included revising the operability evaluations.

Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation is treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000382/2014005-02, Failure to Follow the Operability Determination Process.

1R18 Plant Modifications

a. Inspection Scope

On October 30, 2014, the inspectors reviewed a temporary plant modifications related to the emergency diesel fuel oil feed tank vent lines.

The inspectors verified that the licensee had installed this temporary modification in accordance with technically adequate design documents. The inspectors verified that this modification did not adversely impact the operability or availability of affected SSCs.

The inspectors reviewed design documentation and plant procedures affected by the modification to verify the licensee maintained configuration control.

These activities constitute completion of one sample of temporary modifications, as defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed three post-maintenance testing activities that affected risk-significant structures, systems, or components (SSCs):

• On November 4, 2014, dry cooling tower B

• On November 13, 2014, timing relays associated with the train A emergency diesel generator sequencer

• On December 1, 2014, essential chiller B 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 three post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

.1 Failure to Establish an Inspection Schedule of the Dry Cooling Towers

Introduction.

The inspectors reviewed a self-revealing, Green, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the failure of the licensee to develop a preventative maintenance schedule for the inspection of safety-related equipment. Specifically, the licensee did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers.

Description.

On January 31, 2014, the licensee discovered a leak of approximately 2.7 mL/min at tube bundle 1 of the B train dry cooling tower (DCT). The DCTs provide cooling for the safety-related component cooling water system. The leak originated near the tube-to-tube sheet interface. The licensee took efforts to locate the source of the leak, including performing boroscope inspections and chemistry sampling. The leak ceased after February 17, 2014, before the licensee found the location of the leak. Due to its intermittent nature, the exact location of the leak could not be determined. As a result, the licensee stopped taking actions to address the leak. On February 25, 2014, based on discussions with the DCT vendor, the licensee concluded the leak was due to a mechanical clearance at the tube-to-tube sheet interface since no welding was used to connect the tubes.

On September 18, 2014, the licensee discovered a leak at tube bundle 1 of the B train DCT near the tube-to-tube sheet interface. The licensee measured the leak at 175 mL/min compared to an allowed system leakage of 557 mL/min. Using a boroscope, on September 19, 2014, the licensee found the leak coming from a pinhole in the last row of the tube bundle producing a visible spray of water. On November 4, 2014, the licensee repaired the leak by plugging the leaking tube.

As part of an apparent cause evaluation, the licensee determined that external corrosion of the DCT tubes caused the leak. The licensee also found that even though preventive maintenance existed for the easily accessible portions of the DCT, no regularly scheduled preventative maintenance tasks were in place for inspecting or cleaning the base of the DCT tube rows that were near the fans, where the leak occurred.

Additionally, the licensee found that the floor decking near the tube bundles obscured visual examination of the tubes.

In reviewing the issue, the inspectors noted that the licensee inspected the DCTs in accordance with procedure EN-DC-178, System Walkdowns, Revision 7. However, procedure EN-DC-178 only required inspections of accessible external surfaces of systems. It did not, however, require inspection of all portions of the DCTs. In addition, as part of this review, the inspectors found that licensee procedure SEP-HX-WF3-001, Generic Letter 89-13 Heat Exchanger Test Basis, step 5.4.11, effective as of August 6, 2012, required the licensee to perform visual inspections to monitor the DCTs for degradation, including corrosion. The inspectors noted that step 5.2.8 of procedure SEP-HX-WF3-001 states that heat exchangers may be disassembled to evaluate their cleanliness. It also states that digital pictures and videos should be taken in order to allow for comparisons between inspections. Step 5.3.8 of procedure EP-HX-WF3-001 states that if unacceptable fouling is not readily detectable through visual inspection, visual monitoring should not be used.

When the inspectors requested the forms documenting completion of procedure SEP-HX-WF3-001 for the DCTs, the licensee indicated that they had not been performing those inspections because the requirements were no longer applicable. To justify this stance, the licensee provided engineering request ER-W3-2001-1125-000, which documented the licensees decision in 2001 to use less rigorous testing for the component cooling water system. Since the DCTs are part of the component cooling water system, they were also impacted by the less rigorous testing schedule.

Engineering request ER-W3-2001-1125-000 stated that past degradation problems had been remedied and that the chemistry control program alone was adequate to prevent further degradation of the CCW system. This, combined with procedure EN-DC-178 not requiring inspection of all portions of the DCT, caused several portions of the DCTs to go un-inspected for at least 13 years.

When the licensee ceased monitoring the DCTs under the Generic Letter 89-13 program, they did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers. The licensee did not identify the lack of inspections when the DCTs were removed from the Generic Letter 89-13 program in 2001 or the inclusion of the DCTs in the Generic Letter 89-13 program when procedure SEP-HX-WF3-001 was implemented in 2012. The licensee plans to evaluate the requirements of procedure SEP-HX-WF3-001 as well as their corrective actions.

Analysis.

The inspectors determined that the failure to develop a preventative maintenance schedule for inspections of the dry cooling towers was a performance deficiency. The inspectors determined that this deficiency was reasonably within the licensees ability to foresee and correct. Using the guidance in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors concluded that performance deficiency was more than minor, and therefore a finding, because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to periodically inspect and perform preventive maintenance on portions of the dry cooling towers prevented the licensee from identifying corrosion that eventually degraded the system enough to cause a leak.

Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment.

The inspectors concluded that the finding reflected current licensee performance and had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee did not implement a corrective action program with a low threshold for identifying issues. Specifically, the licensee did not identify this issue when the leak first developed in January 2014 [P.1].

Enforcement.

Technical Specification 6.8.1.a, requires, in part, that procedures shall be established, implemented and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 9.b of Appendix A to Regulatory Guide 1.33, states, in part, that the preventive maintenance schedules should be developed to specify inspection of equipment.

Contrary to the above, between 2001 and 2014, the licensee did not implement a preventative maintenance schedule for inspections of equipment in the component cooling water system. Specifically, the licensee did not perform inspections for all portions of the safety-related dry cooling towers and, as a result, failed to prevent degradation which resulted in a leak on January 31, 2014, and on September 18, 2014.

The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-04930 and CR-WF3-2014-06100. The corrective action taken to restore compliance was to develop preventative maintenance tasks to visually inspect the DCT tubes, including disassembly where necessary. In addition, the DCT tube leak was repaired. Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation was treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000382/2014005-03, Failure to Establish an Inspection Schedule of the Dry Cooling Towers.

.2 Failure to Establish Design Control Measures for the Suitability of Safety-Related Relays

Introduction.

The inspectors reviewed a self-revealing, Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure of the licensee to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not have an adequate replacement frequency for safety-related relays associated with engineered safety features equipment to ensure that all required equipment operated in the time sequence assumed by the safety analysis.

Description.

On October 17, 2013, the licensee initiated condition report CR-WF3-2013-05091 to determine the cause of the inadvertent equipment actuations of safety-related equipment experienced throughout calendar year 2013. Systems affected by the actuations included both trains of the controlled ventilation area system and the shield building ventilation system, train B of the switchgear ventilation system, control room ventilation systems, and charging pump AB. The licensees evaluation determined that the inadvertent actuations were due to the equipment relays spuriously de-energizing.

The de-energization of the relays, which were Allen Bradley relays, caused the associated equipment to start. All of the inadvertent actuations resulted in equipment running in their required, safety-related position.

At the time of this evaluation, there were 33 Allen Bradley relays installed at Waterford 3.

The licensee determined that for four of the 33 relays, spurious de-energization could result in equipment not operating in accordance with its associated safety analysis.

These four relays monitor for under-voltage conditions on the safety-related 4KV emergency busses. The licensee determined that if a loss of offsite power or a safety injection actuation signal were to occur, there would be a 192-second period of time during which a spurious relay de-energization would result in a lockout signal that could prevent equipment from being loaded onto the associated emergency electrical bus.

Equipment loaded onto the emergency bus prior to the de-energization of the relay would continue running. Subsequent equipment loading into the emergency bus would not occur unless operators manually reset the lockout signal. This failure could prevent safety-related equipment from loading and operating within the timeframe assumed by the safety analysis.

On February 2, 2014, the licensee determined that the relay failures were due to relay components degrading faster than initially assumed. Based on vendor information, the licensee had originally established an 18-year replacement frequency for the relays; however, some relays failed as early as 3 years after installation.

The vendor-recommended voltage range for the relays was 96 to 132 VDC. The licensees allowable operational voltage range for the application was 132 to 135 VDC.

Additionally, in the licensees application, the relays would operate at a maximum of 124°F, which is near the high end of the vendors recommended range of -4°F to 140°F.

Prior to installing the Allen Bradley relays, the licensee contacted the vendor about the potential effects of operating the relays at a higher-than-recommended voltage. The licensee also contacted the vendor about the potential effects of operating the relays at a high temperature. However, the licensee did not consider or ask the vendor about the integrated risk of operating the relays at both higher-than-recommended voltage and temperatures in the higher end of the provided band. As a result, the licensee did not perform an evaluation considering both the high voltage and high temperature conditions prior to the installation of the Allen Bradley relays.

To address the condition, the licensee replaced all of the degraded relays. The licensee also determined a more appropriate expected lifetime for the Allen Bradley relays and took action to increase the replacement frequency from 18 years to 3 years.

Analysis.

The failure to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. The inspectors determined that this deficiency was reasonably within the licensees ability to foresee and correct. The inspectors concluded that the performance deficiency was more than minor, and therefore a finding, because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop an adequate replacement frequency for the Allen Bradley relays used to monitor for under-voltage conditions on the safety-related emergency busses could have prevented the equipment from operating in the time sequence assumed by the safety analysis and therefore fail to perform its safety function.

Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined the finding had very low safety significance (Green) because it was a deficiency affecting the qualification of a mitigating system component and the affected equipment maintained its operability.

The inspectors determined the finding reflected current licensee performance and had a cross-cutting aspect in the area of Human Performance, Challenging the Unknown, because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding [H.11].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion III requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.

Contrary to the above, prior to February 2, 2014, the licensee failed to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not establish an adequate replacement frequency or evaluate for suitability of application of safety-related relays operating at voltage and temperature values that were higher than those recommended by the vendor. As a result, safety-related equipment may not have operated in the time sequence assumed by the safety analysis. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-05091. The corrective actions taken to restore compliance included replacing the affected relays and reducing their replacement frequency from 18 years to 3 years.

Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation was treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000382/2014005-04, Failure to Establish Design Control Measures for the Suitability of Safety-Related Relays.

1R22 Surveillance Testing

a. Inspection Scope

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

Reactor coolant system leak detection tests:

• On October 27, 2014, reactor cooling system leak detection Other surveillance tests:

• On October 15, 2014, surveillance testing of HPSI flow control valves

• On October 21, 2014, surveillance testing of emergency feedwater pump AB thermal overload bypass

• On November 7, 2014, surveillance testing of component cooling water pump A 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 SSCs following testing.

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

b. Findings

No findings were identified.

1EP2 Alert and Notification System Evaluation

a. Inspection Scope

The inspectors verified the adequacy of the licensees methods for testing the primary and backup alert and notification system (ANS). The inspectosr interviewed licensee personnel responsible for the maintenance of the primary and backup ANS and reviewed a sample of corrective action system reports written for ANS problems. The inspectors compared the licensees alert and notification system testing program with criteria in NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1; FEMA Report REP-10, Guide for the Evaluation of Alert and Notification Systems for Nuclear Power Plants; and the licensees current FEMA-approved alert and notification system design report, Updated Alert/Notification System Design Report, Revision 7, dated October 2011.

These activities constituted completion of one alert and notification system evaluation sample as defined in Inspection Procedure 71114.02.

b. Findings

No findings were identified.

1EP3 Emergency Response Organization Staffing and Augmentation System

a. Inspection Scope

The inspectors verified the licensees emergency response organization on-shift and augmentation staffing levels were in accordance with the licensees emergency plan commitments. The inspectors reviewed documentation and discussed with licensee staff the operability of primary and backup systems for augmenting the on-shift emergency response staff to verify the adequacy of the licensees methods for staffing emergency response facilities, including the licensees ability to staff pre-planned alternate facilities.

The inspectors also reviewed records of emergency response organization augmentation tests and events to determine whether the licensee had maintained a capability to staff emergency response facilities within emergency plan timeliness commitments.

These activities constitute completion of one emergency response organization staffing and augmentation testing sample as defined in Inspection Procedure 71114.03.

b. Findings

No findings were identified.

1EP5 Maintenance of Emergency Preparedness

a. Inspection Scope

The inspectors reviewed the following for the period August 2012 through October 2014:

• After-action evaluation reports for licensee drills and exercises

• Independent audits and surveillances of the licensees emergency preparedness program

• Self-assessments of the emergency preparedness program conducted by the licensee

• Licensee evaluations of changes made to the emergency plan and emergency plan implementing procedures

• Drill and exercise performance issues entered into the licensees corrective action program

• Emergency preparedness program issues entered into the licensees corrective action program

• Maintenance records for equipment supporting the emergency preparedness program

• Emergency response organization and emergency planner training records The inspectors reviewed summaries of 191 corrective action program entries (condition reports) associated with emergency preparedness, and selected 19 to review against program requirements, to determine the licensees ability to identify, evaluate, and correct problems in accordance with 10 CFR Part 50, Appendix E, IV.F, and planning standard 10 CFR 50.47(b)(14). The inspectors verified that the licensee accurately and appropriately identified and corrected emergency preparedness weaknesses during critiques and assessments.

The inspectors reviewed three licensee evaluations of the impact of changes to the emergency plan and implementing procedure to determine the licensees ability to identify reductions in the effectiveness of the emergency plan in accordance with the requirements of 10 CFR 50.54(q)(3) and 50.54(q)(4). The inspectors verified that evaluations of proposed changes to the licensee emergency plan appropriately identified the impact of the changes prior to being implemented.

These activities constitute completion of one sample of the maintenance of the licensees emergency preparedness program as defined in Inspection Procedure 71114.05.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

a. Inspection Scope

On December 3, 2014, 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 occupational individual and collective radiation exposures as low as is reasonably achievable (ALARA). During the inspection, the inspectors interviewed licensee personnel and reviewed licensee performance in the following areas:

• Site-specific ALARA procedures and collective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements

• ALARA work activity evaluations/post-job reviews, exposure estimates, and exposure mitigation requirements

• The methodology for estimating work activity exposures; the intended dose outcome; the accuracy of dose rate and man-hour estimates; and intended versus actual work activity doses and the reasons for any inconsistencies

• Records detailing the historical trends, current status of tracked plant source terms, and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

• Audits, self-assessments, and corrective action documents related to ALARA planning and controls since the last inspection These activities constitute completion of one sample of occupational ALARA planning and controls as defined in Inspection Procedure 71124.02.

b. Findings

Failure to Adequately Plan and Control Work Activities Related to Alloy 600 Pipe Weld Inspections to Ensure Doses were ALARA

Introduction.

The inspectors identified a finding of very low safety significance (Green)associated with the licensees failure to adequately plan and control work activities related to Alloy 600 ultrasonic examinations during Refueling Outage 19 (RF19).

Description.

While reviewing the post-job review package for Radiological Work Permit (RWP) 2014-0613 from RF19, the inspectors identified that the licensees ALARA planning and control process failed to prevent unplanned and unintended collective doses. Specifically, the original dose estimate for RWP 2014-0613 (Alloy 600 inspections) was 2.75 man-rem, and the actual accumulated collective dose was 7.645 man-rem. This represented an overage of 4.895 man-rem and exceeded the original dose estimate by 178 percent.

The work activities covered under RWP 2014-0613 included preparation and examination of dissimilar metal welds on the reactor coolant pump suction and discharge pipes and the safety injection (SI) nozzles using encoded ultrasonic phased array testing. The inspectors determined that if the licensee had appropriately evaluated the condition and physical layout of each weld during the ALARA planning process and had appropriately performed in-progress ALARA reviews, the significant dose overrun would have been prevented.

During Refueling Outage 16 (RF16), the licensee performed the weld examinations using a manual method instead of the automatic method used in RF19. When planning the level of effort (man-hours) required for these activities, the licensee and the vendor who performed the examinations relied on pictures of the pipe welds from RF16 and a single plane surface profile for each weld. Performing a walkdown, a standard practice prior to nondestructive examinations, would have provided better information about the weld condition, physical layout, and pipe interferences.

The RF16 photographs provided to the vendor did not adequately depict the surface condition of the SI nozzle welds, and, as a result, further unplanned surface preparations had to be performed to adequately prepare the welds. Once the work began, the vendor had to change the type of prep tool used after prepping the first three welds, and spent significantly more time than had been estimated on weld preparation. The original time estimated for weld preparation was 71 man-hours; the actual time required was 418 man-hours.

In addition, the photographs provided by the vendor did not fully document the physical layout of the piping and potential pipe interferences. Based on their past experience at Waterford 3 and the pictures provided, the vendor built a mock-up of the pipe to estimate the hours needed and practice the examinations. However, neither the mock-up nor the pictures detailed the interferences on the pipe, which required the vendor to constantly adjust the rail system on which the examination equipment rides. The failure to identify and plan for physical interferences encountered on the pipe welds resulted in the examinations taking longer than planned. The original time estimate for the examinations was 84 man-hours; the actual time was 624 man-hours.

With respect to ALARA planning for the work, the inspectors concluded that the licensee did not follow Section 5.3, RWP Planning Process, of Procedure EN-RP-105, Radiological Work Permits, Revision 14, which addresses ALARA planning.

Specifically, Step 5 states: WHEN using historical survey OR work information, THEN ensure it is representative of the work area radiological conditions and the work activities to be performed. However, neither the licensee nor the vendor ensured that the historical work information (e.g., photographs of the welds, time to perform the work during RF16) was representative of the work activities to be performed during RF19 by performing walk downs of the areas during the planning stage or prior to initiating the work.

The inspectors also determined that the licensee ALARA process had controls in place that could have foreseen and corrected the inaccurate dose/time estimate once the work began. The licensee did not follow steps of Procedure EN-RP-105 which address in-progress reviews. Specifically, the licensee failed to appropriately complete the following actions:

Section 5.6: RWP In-Progress Reviews o Section 5.6[2]: RWP In-Progress Reviews are normally performed at the following intervals for specific RWPs with a dose estimate equal to or greater than 1 man-rem and less than 5 man-rem

• Approximately 50 percent of dose estimate o Section 5.6[3]: RWP In-Progress Reviews are normally performed at the following intervals for specific RWPs with a dose estimate equal to or greater than 5 man-rem

• Approximately 40 percent of RWP dose estimate

• Approximately 80 percent of RWP dose estimate o Section 5.6[5]: IF during the performance of an In-Progress Review any of the following deficiencies are identified,

• RWP revision is necessary because RWP exceeds 125 percent of the current dose estimate

• Indications of poor planning Then initiate a Condition Report.

The inspectors determined the licensee did not perform an in-progress review upon reaching approximately 50 percent of the dose estimate as stated in the procedure.

Documentation reviewed stated the licensee performed an in-progress review at 80 percent of the dose estimate; however, the actual dose at the time of review was 116 percent of the original dose estimate. The in-progress review did not address why neither a 40 percent nor a 50 percent review was performed, as stated in their procedure. Furthermore, when the dose exceeded 125 percent of the original dose estimate, a condition report was not initiated.

The two in-progress reviews that were performed (after reaching 116 percent of the original dose estimate) did not describe any changes to radiological conditions or fully evaluate the cause of the dose overruns. In each case, the licensee added dose to the estimate, pointing to the inaccurate time estimate for the work activities, but did not address the cause for the additional time nor propose ways to reduce dose accumulation. The inspectors determined that the licensee did not document a sound justification for adding 1.5 man-rem to the RWP in Revision 2 or adding 5.136 man-rem to the RWP in Revision 3.

The inspectors determined that if the licensee had appropriately evaluated the Alloy 600 pipe weld conditions/locations during the ALARA planning process and appropriately performed in-progress ALARA reviews, then they could have reasonably planned for the full scope of work and provided a better estimate, and/or adequately justified revising the estimate, for the job. These failures to plan and control the job activities led to unplanned, unintended collective dose. The finding and procedure concerns were documented in the licensees corrective action program as CR-WF3-2014-06343.

Analysis.

The failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations is a performance deficiency. This performance deficiency is more than minor because it is associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the decision to rely on photographs and not walk down the pipe welds to determine their physical condition and location, and the failure to perform timely in-progress ALARA reviews, caused the collective radiation dose to be greater than 5 man-rem and exceed the planned dose estimate by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding has very low safety significance (Green) because:

(1) it was associated with ALARA planning and work controls and

(2) the licensees three-year rolling average collective dose of 121.7 man-rem, which placed them at the bottom of the 4th quartile among PWRs, was less than 135 man-rem.

The finding has a Work Management cross-cutting aspect, associated with the Human Performance cross-cutting area, because the licensee did not adequately plan or control work activities such that nuclear safety is the overriding safety priority. Specifically, the ALARA plan did not reflect the time needed to complete the work activities, thus underestimating the dose requirements, and the administrative control of reviewing the work-in-progress at appropriate completion points failed (H.5).

Enforcement.

This finding does not involve enforcement action because no violation of regulatory requirements was identified. However, the performance deficiency is directly related to the licensees failure to meet the expectation to fully implement Procedure EN-RP-105, Radiological Work Permits, Revision 14, Sections 3.0, 5.3, and 5.6. The finding is documented in the licensees corrective action program by Condition Report CR-WF3-2014-06343. Because this finding does not involve a violation and is of very low safety significance, it is identified as a FIN 05000382/2014005-05, Failure to adequately plan and control work activities related to Alloy 600 pipe weld inspections to ensure doses were ALARA.

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 one sample 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 Drill/Exercise Performance (EP01)

a. Inspection Scope

The inspectors reviewed the licensees evaluated exercises, emergency plan implementations, and selected drill and training evolutions that occurred between October 2013 and September 2014 to verify the accuracy of the licensees data for classification, notification, and protective action recommendation (PAR) opportunities.

The inspectors reviewed a sample of the licensees completed classifications, notifications, and PARs to verify their timeliness and accuracy. The inspectors used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.

These activities constituted verification of the drill/exercise performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Emergency Response Organization Drill Participation (EP02)

a. Inspection Scope

The inspectors reviewed the licensees records for participation in drill and training evolutions reported between October 2013 and September 2014 to verify the accuracy of the licensees data for drill participation opportunities. The inspector verified that all members of the licensees emergency response organization (ERO) in the identified key positions had been counted in the reported performance indicator data and reviewed the licensees basis for reporting the percentage of ERO members who participated in a drill.

The inspectors reviewed drill attendance records and verified a sample of those reported as participating. The inspector used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.

These activities constituted verification of the emergency response organization drill participation performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Alert and Notification System Reliability (EP03)

a. Inspection Scope

The inspectors reviewed the licensees records of Alert and Notification System tests conducted between October 2013 and September 2014 to verify the accuracy of the licensees data for siren system testing opportunities. The inspectors reviewed procedural guidance on assessing Alert and Notification System opportunities and the results of periodic alert and notification system operability tests. The inspectors used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.

These activities constituted verification of the alert and notification system reliability performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.4 Reactor Coolant System Specific Activity (BI01)

a. Inspection Scope

The inspectors reviewed the licensees reactor coolant system chemistry sample analyses for the period of January 2013 through September 2014 to verify the accuracy and completeness of the reported data. The inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample on December 16, 2014. 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 reactor coolant system specific activity performance indicator for Unit 3, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.5 Reactor Coolant System Identified Leakage (BI02)

a. Inspection Scope

The inspectors reviewed the licensees records of reactor coolant system identified leakage for the period of January 2013 through September 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 reactor coolant system leakage performance indicator for Unit 3, 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 list and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors focused their review on the licensees timelines for the generation of condition reports following the identification of a condition adverse to quality. 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. Observations and Assessments The inspectors identified a trend with respect to the licensees timeliness when generating condition reports. Specifically, for the six month trend inspection period the inspectors identified several examples where a condition adverse to quality was identified, either by the inspectors or licensee personnel, but a condition report was not generated in a timely manner. Examples of the conditions adverse to quality included increased vibrations in a turbine bearing, vital and fire doors left open and unattended, degraded dry cooling tower tubes, and operability reviews on safety-related relays. The inspectors noted that in most instances licensee personnel would delay the generation of a condition report until they were able to collect additional information to document the condition. This resulted in extended periods of time before the condition was entered into the corrective action program. In other instances, the inspectors noted a lack of consistency with respect to the threshold to generate a condition report. Specifically, the inspectors identified examples where personnel that were aware of a condition adverse to quality would first try to resolve the issue without entering it into the corrective action program or they were not aware that they had to generate a condition report.

The inspectors communicated this trend to the licensee. The licensee initiated an apparent cause evaluation to evaluate the causes for this adverse trend.

c. Findings

Failure to Correct a Condition Adverse to Quality in a Timely Manner

Introduction.

The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a condition adverse to quality in a time commensurate with the safety significance of the issue. Specifically, the licensee failed to repair degraded conduit that had been identified as corroded since 2008. As a result, conduits that were housing cables for safety-related components were degraded to the point where water entered the conduit and submerged cables that were not designed for submergence for an extended period of time.

Description.

During a routine review of items entered in the licensees Corrective Action Program (CAP), the inspectors identified a CAP item documenting that a conduit was found to have water dripping out of it in the +35 foot elevation of the Reactor Auxiliary Building, inside the Cable Spreading Room. The CAP item documented that the conduit that was dripping water originated at the +46 foot elevation of the Reactor Auxiliary Building in the East Wing Area. The East Wing Area, an area exposed to the outside environment, is located directly above the Cable Spreading Room. The CAP item noted that the conduit in the +46 foot elevation East Wing Area was severely corroded to the point that it was breached, and rain water entered and filled the conduit. Because the conduit penetrates the floor of the East Wing Area and travels down to the Cable Spreading Room, the water that had accumulated inside the conduit had also corroded part of the conduit inside the Cable Spreading Room. This resulted in the leaking water that was initially identified.

The inspectors toured the areas mentioned in the CAP item and noticed that there were other conduits carrying safety-related cables inside the Cable Spreading Room that were also corroded and leaking water. Some of the safety-related cables that were impacted by the water-filled conduits were cables for the atmospheric dump valve #2, main steam isolation valve #2 upstream drip pot startup drain, main steam isolation valve #2 upstream drip pot normal drain, and nitrogen accumulator #7 stop valve which provides air for valves in the emergency feed water system. The inspectors reviewed the associated items in the CAP and noticed that there were CAP items dating back to 2008 that documented the condition of corroded conduits. When the inspectors evaluated the licensees corrective actions for these CAP items, they noted that the degraded conduit condition was first identified in condition report CR-WF3-2008-03646. That condition report was closed to a work order which was still open in 2014 when the issue was documented in the CAP again in condition report CR-WF3-2014-03366. In addition, there were other condition reports that were generated in 2011, documenting the same condition (degraded conduits in the East Wing Area and the Cable Spreading Room),that were also left open and uncorrected.

The inspectors also noted that the safety-related cables that had been inside the flooded conduits since at least 2008 were qualified for above ground and underground applications in wet and dry locations. The cables were not rated for submergence.

Having the cables submerged for extended periods of time could result in sufficient degradation to impact the function of the associated components in the event of an accident. The inspectors determined that the licensee had opportunities dating back to 2008 to repair the degraded conduit such that safety-related cables that are not rated for submergence would not be submerged in water for extended periods of time, thereby potentially impacting the operability and availability of safety-related components.

Analysis.

The inspectors determined that the licensees failure to correct a condition adverse to quality in a timely manner was a performance deficiency. Using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the performance deficiency was more than minor, and therefore a finding, because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, safety-related cables that were not rated for full submergence were submerged in water since at least 2008, thereby degrading the integrity of the cable and potentially impacting the ability of safety-related equipment to perform safety functions in the event of an accident.

The inspectors used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power, to determine the significance of the findings. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding screened as having very low safety significance (Green)because the finding impacted the qualification of mitigating components, but the components maintained operability.

This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensees decision-making practices did not emphasize prudent choices over those that are simply allowable. Specifically, when evaluating the condition reports written through the years that documented the degraded conduits, the licensee elected to defer needed maintenance instead of placing the adequate priority on the issue. [H.14]

Enforcement.

Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, states in part that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconfrmances are promptly identified and corrected.

Contrary to the above, between August 2008 and June 2014, the licensee failed to promptly correct a condition adverse to quality. Specifically, after identifying in August 2008, that safety-related cables in the East Wing Area and Cable Spreading Room were submerged in water due to degraded conduits, the licensee failed to take corrective action to correct the condition in a timely manner. As a result, the degraded condition went uncorrected for 6 years, and it could have impacted the availability of safety-related equipment in the event of an accident. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04951. The corrective actions taken to restore compliance included performing an extent of condition review to identify other cables that could potentially be impacted by degraded conduits, and repairing the degraded conduit for the impacted safety-related cables.

Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation was treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000382/2014005-06; Failure to Correct a Condition Adverse to Quality in a Timely Manner.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On November 21, 2014, the emergency preparedness inspector presented the results of the on-site inspection of the licensees emergency preparedness program 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 December 16, 2014, the radiation safety 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 January 15, 2015, the resident inspectors presented the 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.

4OA7 Licensee-Identified Violations

The following Severity Level IV violation was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.

• Title 10 of the Code of Federal Regulations, Appendix E to Part 50,Section V, states that licensees who are authorized to operate a nuclear power facility shall submit any changes to the emergency plan or procedures to the Commission, as specified in 50.4, within 30 days of such changes. Title 10 of the Code of Federal Regulations, Section 50.54(q)(5) states, in part, that licensees shall submit a report of changes made after February 21, 2012, that includes a summary of its analysis, within 30 days after the change is put into effect. Contrary to the above, Waterford 3 Steam Electric Station did not submit changes to emergency plan implementing procedures within 30 days of such changes, and did not submit a summary of its analysis of the changes within 30 days after the changes were put into effect. Specifically, the licensee did not submit changes to the following procedures: EN-EP-305, Emergency Planning 10CFR50.54(Q)

Review Program, Revision 3; EN-EP-306, Drills and Exercises, Revisions 4 and 5; EN-EP-308, Emergency Planning Critiques, Revision 2; EN-EP-310, Emergency Response Organization Notification System, Revisions 1 through 3; EN-EP-311, Emergency Response Data System (ERDS) Activation Via the Virtual Private Network (VPN), Revision 2; EN-EP-313; Offsite Dose Assessment Using the Unified RASCAL Interface, Revision 0; EN-EP-801, Emergency Response Organization, Revision 8; EN-TQ-110, Emergency Response Organization Training, Revision 7, and EN-TQ-110-01, Fleet E-Plan Training Course Summary, Revision 10.

The licensee did not have a process to ensure that fleet procedures necessary to implement the site emergency plan were submitted to the NRC in accordance with the requirements of Appendix E to 10 CFR Part 50. This violation was evaluated using the NRC Enforcement Policy because the licensees failure to submit required procedures affected the NRCs ability to perform adequate regulatory oversight, and was evaluated as a Severity Level IV violation because the violation was not related to the licensees ability to perform notification or assessment during an emergency. This issue has been entered into the licensees corrective action program as Condition Reports CR-HQN-2014-00380, CR-HQN-2014-00597, and CR-WF3-2014-05727.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Chisum, Site Vice President, Operations

M. Richey, General Manager, Plant Operations

J. Brawley, Supervisor, Radiation Protection

J. Briggs, Superintendent, Electrical Maintenance

L. Brown, Licensing Specialist, Regulatory Assurance

M. Chaisson, Supervisor, Radiation Protection

K. Crissman, Senior Manager, Maintenance

D. Frey, Manager, Radiation Protection

R. Gilmore, Manager, Systems and Components

A. James, Manager, Security

J. Jarrell, Manager, Regulatory Assurance

B. Lanka, Director, Engineering

N. Lawless Manager, Chemistry

B. Lindsey, Senior Manager, Operations

R. Osborne, Manager, Performance Improvement

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

M. Mills, Manager, Nuclear Oversight

L. Milster, Licensing Specialist, Licensing

B. Pellegrin, Senior Manager, Production

J. Pollock, Licensing Specialist, Licensing

R. Porter, Manager, Design & Program Engineering

D. Reider, Supervisor, Quality Assurance

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

W. McKinney, Acting Manager, Training

R. Simpson, Superintendent, Operator Training

N. Petit, Supervisor, Design Engineering

M. Vierra, Dosimetry Technician, Radiation Protection

J. Vollmer, CHP, Specialist, Radiation Protection

J. Williams, Senior Licensing Specialist

NRC Personnel

F. Ramírez, Senior Resident Inspector

C. Speer, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

Failure to Identify and Control Potential Tornado-Borne

05000382-2014005-01 NCV Missile Hazards (Section 1R01)

Failure to Follow the Operability Determination Process

05000382-2014005-02 NCV when (Section 1R15)

Attachment 1

Opened and Closed

Failure to Establish an Inspection Schedule of the Dry

05000382-2014005-03 NCV Cooling Towers (Section 1R19)

Failure to Establish Design Control Measures for the

05000382-2014005-04 NCV Suitability of Safety-Related Relays (Section 1R19)

Failure to Adequately Plan and Control Work Activities

05000382-2014005-05 FIN Related to Alloy 600 Pipe Weld Inspections to Ensure Doses were ALARA (Section 2RS2)

Failure to Correct a Condition Adverse to Quality in a

05000382-2014005-06 NCV Timely Manner (Section 4AO2)