IR 05000390/2013004

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IR 05000390-13-004; on 07/01/2013 - 09/30/2013; Watts Bar, Unit 1; Integrated Report
ML13309B280
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 11/05/2013
From: Scott Shaeffer
Reactor Projects Region 2 Branch 6
To: James Shea
Tennessee Valley Authority
References
IR-13-004
Download: ML13309B280 (35)
v • d • e


Text

UNITED STATES ber 5, 2013

SUBJECT:

WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000390/2013004

Dear Mr. Shea:

On September 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Unit 1. The enclosed inspection report documents the inspection results which were discussed on October 2, 2013, with members of the Watts Bar staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

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

No NRC-identified findings or Self-Revealing findings are documented.

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

Sincerely,

/RA/

Scott M. Shaeffer, Chief Reactor Projects Branch 6 Division of Reactor Projects Docket No.: 50-390 License No.: NPF-90

Enclosure:

NRC Inspection Report 05000390/2013004 w/Attachment: Supplemental Information

REGION II==

Docket No.: 50-390 License No.: NPF-90 Report No.: 05000390/2013004 Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Unit 1 Location: Spring City, TN 37381 Dates: July 1 through September 30, 2013 Inspectors: R. Monk, Senior Resident Inspector K. Miller, Resident Inspector R. Hamilton, Senior Health Physicist (Sections 2RS2, 2RS4, 4OA1)

A. Nielsen, Senior Health Physicist (Sections 2RS1, 4OA1)

R. Kellner, Health Physicist (Sections 2RS3, 2RS5)

Approved by: Scott M. Shaeffer, Chief Reactor Projects Branch 6 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000390/2013-004; 07/01/2013 - 09/30/2013; Watts Bar, Unit 1; Integrated Report.

The report covered a three-month period of inspection by resident inspectors and announced inspections by regional inspectors. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 4, dated December 2006.

NRC-Identified Findings and Self-Revealing Findings

None

Licensee-Identified Violations

None

REPORT DETAILS

Summary of Plant Status

Unit 1 was in the process of ramping up in power from a June 28, 2013 reactor trip. It achieved 100% rated thermal power and has operated at essentially 100% until the end of the report period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 External Flooding

a. Inspection Scope

Inspectors observed a detailed plant exercise of AOI-7.01, Maximum Probable Flood.

This included staffing of the Central Emergency Control Center, CECC in Chattanooga, the Technical Support Center, TSC, a mock control room and the Operations Support Center, OSC. Using appropriate time compression, the licensee demonstrated to a plausible degree that changes incorporated into AOI-7.01 were adequate to accomplish the complex activities to place the plant into the flood mode configuration to protect the reactor core during a design basis Probable Maximum Flood.

b. Findings

No findings were identified.

1R04 Equipment Alignment

Partial System Walkdowns

a. Inspection Scope

The inspectors conducted three equipment alignment partial walkdowns, listed below, to evaluate the operability of selected redundant trains or backup systems with the other train or system inoperable or out of service. This also includes that redundant trains are returned to service properly. The inspectors reviewed the functional system descriptions, Updated Final Safety Analysis Report (UFSAR), system operating procedures, and Technical Specifications (TS) to determine correct system lineups for the current plant conditions. The inspectors performed walkdowns of the systems to verify that critical components were properly aligned and to identify any discrepancies which could affect operability of the redundant train or backup system. Documetns reviewed are listed in the Attachment.

  • Auxiliary control air system A-train while B-train OOS for maintenance (SOI-32.02)

.2 Complete System Walkdown

a. Inspection Scope

The inspectors conducted one detailed walkdown/review of the alignment and condition of the ERCW to the auxiliary building components to verify proper equipment alignment and to identify any discrepancies that could impact the function of the system and increase risk. The inspectors utilized licensee procedures, as well as licensing and design documents, when verifying that the system alignment was correct. During the walkdown, the inspectors also verified, as appropriate, that: 1) valves were correctly positioned and did not exhibit leakage that would impact the function(s) of any valve; 2) electrical power was available as required; 3) major portions of the system and components were correctly labeled, cooled, ventilated, etc.; 4) hangers and supports were correctly installed and functional; 5) essential support systems were operational; 6) ancillary equipment or debris did not interfere with system performance; 7) tagging clearances were appropriate; and 8) valves were locked as required by the licensees locked valve program. Pending design and equipment issues were reviewed to determine if the identified deficiencies significantly impacted the systems functions.

Items included in this review were the operator workaround list, the temporary modification list, system health reports, and outstanding maintenance work requests/work orders (WOs). In addition, the inspectors reviewed the licensees corrective action program (CAP) to ensure that the licensee was identifying equipment alignment problems and to ensure they were properly addressed for resolution.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope

The inspectors conducted tours of the six areas important to reactor safety, listed below, to verify the licensees implementation of fire protection requirements as described in the Fire Protection Program, Nuclear Power Group Standard Programs and Processes (NPG-SPP)-18.4.6, Control of Fire Protection Impairments, NPG-SPP-18.4.7, Control of Transient Combustibles, NPG-SPP-18.4.8, Control of Ignition Sources (Hot Work). The inspectors evaluated, as appropriate, conditions related to: 1) licensee control of transient combustibles and ignition sources; 2) the material condition, operational status, and operational lineup of fire protection systems, equipment, and features; and 3) the fire barriers used to prevent fire damage or fire propagation. This activity constituted six inspection samples.

  • Auxiliary instrument room
  • Intake pumping station

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

The inspectors performed one heat sink performance review. The inspectors reviewed the licensees program for maintenance and testing of two risk-important heat exchangers in the ERCW system. Specifically, the review included the program for testing and analysis of the A (WBN-1-HTX-070-0185) and C (WBN-0-HTX-070-0186)component cooling system (CCS) heat exchangers. The inspectors reviewed the ERCW system description and the heat exchanger testing program document as well as completed WOs documenting the testing and visual inspection and associated corrective actions to verify that corrosion or fouling did not impact the heat exchanger from achieving its design basis heat removal capacity. The inspectors also reviewed periodic test data of ERCW and CCS flow rates as well as inlet and outlet temperatures to determine whether potential degradations were being monitored and/or prevented.

Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

Routine Operator Requalification Review: On August 28, 2013, the inspectors observed the simulator evaluations for Operations Crew 3 per 3-OT-SRE0052, Rev. 0, EHC leak with Large LOCA and 3-OT-SRE0005, Rev 0, Main Steam Line Break Inside Containment/SGTR. The plant conditions led to an Alert level classification for each scenario. Performance indicator credit was taken.

The inspectors specifically evaluated the following attributes related to the operating crews performance:

  • Clarity and formality of communication
  • Ability to take timely action to safely control the unit
  • Prioritization, interpretation, and verification of alarms
  • Correct use and implementation of abnormal operating instructions and emergency operating instructions
  • Timely and appropriate Emergency Action Level declarations per emergency plan implementing procedures
  • Control board operation and manipulation, including high-risk operator actions
  • Command and Control provided by the unit supervisor and shift manager The inspectors also attended the critique to assess the effectiveness of the licensee evaluators and to verify that licensee-identified issues were comparable to issues identified by the inspector.

Observation of Operator Performance: Inspectors observed and assessed licensed operator performance in the plant and main control room, particularly during periods of heightened activity or risk and where the activities could affect plant safety. Inspectors reviewed various licensee policies and procedures such as procedures OPDP-1, Conduct of Operations; NPG-SPP-10.0, Plant Operations; and GO-4, Normal Power Operation.

Inspectors utilized activities such as post maintenance testing, surveillance testing and refueling, and other outage activities to focus on the following conduct of operations as appropriate:

  • Operator compliance and use of procedures
  • Control board manipulations
  • Communication between crew members
  • Use and interpretation of plant instruments, indications and alarms
  • Use of human error prevention techniques
  • Documentation of activities, including initials and sign-offs in procedures
  • Supervision of activities, including risk and reactivity management
  • Pre-job briefs

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the two performance-based problems listed below. A review was performed to assess the effectiveness of maintenance efforts that apply to scoped structures, systems, or components (SSCs) and to verify that the licensee was following the requirements of TI-119, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting 10 CFR 50.65, and NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting 10 CFR 50.65. Reviews focused, as appropriate, on: 1) appropriate work practices; 2) identification and resolution of common cause failures; 3) scoping in accordance with 10 CFR 50.65; 4) characterization of reliability issues; 5) charging unavailability time; 6) trending key parameters; 7) 10 CFR 50.65 (a)(1) or (a)(2) classification and reclassification; and 8) the appropriateness of performance criteria for SSCs classified as (a)(2) or goals and corrective actions for SSCs classified as (a)(1).

  • Removal of hydrogen recombiners from category a(1)
  • Review of Rev. 1 of corrective action plan for CCS containment isolation check valves

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors evaluated, as appropriate, for the three work activities listed below:

1) the effectiveness of the risk assessments performed before maintenance activities were conducted; 2) the management of risk; 3) that, upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and 4) that maintenance risk assessments and emergent work problems were adequately identified and resolved. The inspectors verified that the licensee was complying with the requirements of 10 CFR 50.65 (a)(4); NPG-SPP-07.0, Work Control and Outage Management; NPG-SPP-07.1, On Line Work Management; and TI-124, Equipment to Plant Risk Matrix. This inspection satisfied three inspection samples for Maintenance Risk Assessment and Emergent Work Control.

  • Risk assessment for work week 110 with 1A SFP pump and heat exchanger OOS (yellow) and 2A EDG OOS for planned maintenance
  • Risk assessment for work week 111 with 2B CCS pump and 1B EDG OOS for planned maintenance
  • Risk assessment for work week 112 with 1B safety injection pump and turbine-driven auxiliary feedwater pump OOS for planned maintenance

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed five operability evaluations affecting risk-significant mitigating systems, listed below, to assess, as appropriate: 1) the technical adequacy of the evaluations; 2) whether continued system operability was warranted; 3) whether the compensatory measures, if involved, were in place, would work as intended, and were appropriately controlled; 4) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation (LCOs) and the risk significance in accordance with the significant determination process (SDP). The inspectors verified that the operability evaluations were performed in accordance with NPG-SPP-03.1, Corrective Action Program.

  • Prompt determination of operability (PDO) for PER 756835, Commercial dedication process for replacement shaft used in B-train shutdown board room chilled water pump
  • PDO for PER 759524, 125vdc vital power system response to a loss of coolant event without a loss of offsite power event
  • PDO for PER 756609 Fire wrap conduit IPP3026A installed with threaded fitting, may leak during High Energy/Moderate Energy Line Break (HELB/MELB)
  • PDO for PER 751896, Elastimold separable connectors for
(4) ERCW pump cables do not meet flood mode requirements

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed one permanent plant modification to verify that design output controls were adequate, post-modification testing was satisfactory, and affected operational procedures and licensing documents were identified and revised accordingly. Documents reviewed are listed in the Attachment.

  • Design Change Notice (DCN) 58784A-STG11, Modification of ERCW Strainer 1A-A Backwash Valve 1-FCV-9A-A

b. Findings

No findings were identified.

.2 (Closed) Unresolved Item (URI)05000390/2012005-03, Engineering Justification for

Modifications to Non-Conforming Baskets

a. Inspection Scope

The Watts Bar 1 Integrated inspection report 05000390/2012005 identified a URI associated with the engineering justification for the use of an engineering document change (EDC) and an equivalent change (EQV) for hardware modifications to ice baskets that were nonconforming. The inspectors reviewed outage WO 113393057 which specified the installation of new hardware on a total of six ice baskets that had been damaged, apparently due to ice condenser maintenance. Each of the six damaged baskets (non-conforming components) had suffered plastic deformation (local compressive buckling) of support ligaments in the vicinity of the bottom three feet of the baskets. Instead of replacing the damaged portion of the baskets, as permitted by FSAR Section 6.7.4, licensee engineering had designed hardware to add to the damaged portions of the baskets. Document EDC E-50607, Revision A, specified the installation of vertical supports mechanically attached on the outside of the damaged area and EQV 60275, Revision A, specified the use of wire rope (steel cable) laced through the damaged area. According to the referenced calculation, WCG-1-1912, Qualification of the Optional Lower Ice Basket Support, the vertical supports were intended for compressive loading and the wire rope was intended for tensile loading.

Per FSAR Table 6.7-2, during a deadweight load or deadweight and seismic loads the vertical load on the ice baskets is in compression. When subjected to a design basis accident (DBA) load in combination with a deadweight, or deadweight and earthquake load, the vertical load on all the ice baskets is in tension and the compressed ice basket would tend to elongate.

Review of FSAR Section 6.7.4.3, Design Evaluation, Loading Conditions, part 2.,

Blowdown Loads, subpart E,. Horizontal Ice Basket Forces, states that the tangential and radial forces acting on the ice baskets due to cross-flow are assumed to act on the bottom three feet of ice basket (one-half of the span between the top of the lower support structure and the attachment of the ice baskets to the first lattice frame). The inspectors questioned if the licensee-developed design changes adequately considered these dynamic tangential and radial loads on the damaged ice baskets. The inspectors were concerned that the modifications may not be adequate for the damaged ice baskets to withstand all static and dynamic loads they were originally designed, tested, and qualified to be subjected to. Although there was verbal contact between the licensee and the Original Equipment Manufacturer (OEM) engineering organization regarding the damaged ice baskets, there was no formal written OEM review and acceptance of the licensee modifications as an acceptable alternative to ice basket replacement or repair per FSAR Section 6.7.4.

The licensee subsequently contracted with the OEM engineering organization regarding the evaluation of repairs to the six damaged ice baskets. The OEM responded with a written report that addressed all of the inspectors design concerns documented in the URI. The OEM engineering organization confirmed that the licensee had adequately addressed the appropriate design basis loads in their repair documents. The OEM stated that circumferential (tangential) flow provides a negligible effect that in the case of TVAs (licensees) repair documents do not need to be addressed separately in addition to the radial and vertical flows. The OEM confirmed that the installation of vertical supports mechanically attached on the outside of the damaged area were comparable in design to a proposed OEM repair methodology presented to licensees on or about August 22, 1998. The OEM engineers stated that they considered the subject licensee repairs to be permanent. Thus, it would not be necessary to replace the damaged/repaired portions of these six damaged ice condenser baskets. Based upon review and verification of this information for the hardware modifications to six damaged ice baskets, the inspectors confirmed that the repairs were adequate, as documented in the OEM engineering evaluation.

b. Findings

No findings were identified. This URI is closed.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed three post-maintenance test procedures and/or test activities, (listed below) as appropriate, for selected risk-significant mitigating systems to assess whether: 1) the effect of testing on the plant had been adequately addressed by control room and/or engineering personnel; 2) testing was adequate for the maintenance performed; 3) acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing basis documents; 4) test instrumentation had current calibrations, range, and accuracy consistent with the application; 5) tests were performed as written with applicable prerequisites satisfied; 6) jumpers installed or leads lifted were properly controlled; 7) test equipment was removed following testing; and 8) equipment was returned to the status required to perform its safety function. The inspectors verified that these activities were performed in accordance with NPG-SPP-06.9, Testing Programs; NPG-SPP-06.3, Pre-/Post-Maintenance Testing; and NPG-SPP-07.1, On Line Work Management.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors witnessed five surveillance tests and/or reviewed test data of selected risk-significant SSCs, listed below, to assess, as appropriate, whether the SSCs met the requirements of the TS; the UFSAR; NPG-SPP-06.9, Testing Programs; NPG-SPP-06.9.2, Surveillance Test Program; and NPG-SPP-09.1, ASME Section XI. The inspectors also determined whether the testing effectively demonstrated that the SSCs were operationally ready and capable of performing their intended safety functions.

In-Service Test:

  • WO 114667585, 1-SI-85-2, Reactivity control systems movable control assemblies (Modes 1 and 2) quarterly surveillance test

b. Findings

No findings were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation

a. Inspection Scope

On September 4, 2013, the inspectors observed a licensee-evaluated emergency preparedness drill, listed below, to verify that the emergency response organization was properly classifying the event in accordance with procedure EPIP-1, Emergency Plan Classification Flowchart, and making accurate and timely notifications and protective action recommendations in accordance with EPIP-2, Notification of Unusual Event; EPIP-3, Alert; EIPIP-4, Site Area Emergency; EPIP-5, General Emergency; and the Radiological Emergency Plan. In addition, the inspectors verified that licensee evaluators were identifying deficiencies and properly dispositioning performance against the performance indicator criteria in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline.

  • Waste gas decay tank release below NOUE level.
  • #2 RCP develops a locked rotor condition with the breaker failing to trip, resulting in a partial loss of off-site power.
  • #2 RCP seal leakage results in a safety injection and EAL classification 1.2.2P Non Isolable RCS leak exceeding the capacity of one charging pump.
  • 1A EDG fails and 1B CS pump fails to inject resulting in EAL 1.3.2P Pressure > 2.8 psig (Phase B) with < one full train of containment spray.
  • Fuel failure and resulting radiation monitor reading result in a General Emergency declaration.
  • Appropriate protective action recommendations (PARS) made.

b. Findings

No findings were identified

RADIATION SAFETY

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Hazard Assessment and Instructions to workers During facility tours, the inspectors directly observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRAs), and airborne radioactivity areas established within the radiologically controlled area (RCA) of the auxiliary building and radioactive waste (radwaste) processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for selected RCA areas. The inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, gamma surveys with a range of dose rate gradients, and pre-job surveys for upcoming tasks. The inspectors also discussed changes to plant operations that could contribute to changing radiological conditions since the last inspection. For selected at-power containment entries, the inspectors attended pre-job briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers.

Hazard Control and Work Practices The inspectors evaluated access barrier effectiveness for selected Locked High Radiation Area (LHRA) locations and discussed changes to procedural guidance for LHRA and Very High Radiation Area (VHRA)controls with health physics (HP) supervisors. The inspectors observed and evaluated controls for the storage of irradiated material within the spent fuel pool (SFP).

Established radiological controls (including airborne controls) were evaluated for selected at-power entries into containment and for maintenance work in the SFP transfer canal. In addition, the inspectors reviewed and discussed licensee controls for areas where dose rates could change significantly as a result of incore drive maintenance.

Through direct observations and interviews with licensee staff, the inspectors evaluated occupational workers adherence to selected RWPs and HP technician (HPT) proficiency in providing job coverage. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results for selected at-power entries into containment and for transfer canal maintenance work. The inspectors reviewed the use of personnel dosimetry (extremity dosimetry and multibadging in high dose rate gradients) for the transfer canal maintenance work. The inspectors also evaluated worker response to dose and dose rate alarms during selected work activities.

Control of Radioactive Material The inspectors observed surveys of material and personnel being released from the RCA using small article monitor, personnel contamination monitor, and portal monitor instruments. As part of Inspection Procedure (IP) 71124.05, the inspectors reviewed the last two calibration records for selected release point survey instruments and discussed equipment sensitivity, alarm setpoints, and release program guidance with licensee staff. Also as part of IP 71124.05, the inspectors compared recent 10 Code of Federal Regulations (CFR) Part 61 results for the Dry Active Waste (DAW) radioactive waste stream with radionuclides used in calibration sources to evaluate the appropriateness and accuracy of release survey instrumentation. The inspectors reviewed records of leak tests on selected sealed sources and discussed nationally tracked source transactions with licensee staff.

Problem Identification and Resolution The inspectors reviewed Corrective Action Program (CAP) documents associated with radiological hazard assessment and exposure control. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedures. The inspectors also reviewed recent self-assessment results.

The inspectors evaluated radiation protection activities against the requirements and guidance of Updated Final Safety Analysis Report (UFSAR) Section 12; Technical Specifications (TS) Section 5.11; 10 CFR Parts 19 and 20; Regulatory Guide (RG) 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS2 As Low As Reasonably Achievable (ALARA)

a. Inspection Scope

Work Planning and Exposure Tracking The inspectors reviewed work activities and their collective exposure estimates for the Watts Bar Unit 1 Refueling Outage 11 (1RFO11)and the ongoing transfer canal work being done on Unit 2. The inspectors reviewed ALARA planning packages for the following high collective exposure tasks: temporary shielding, scaffolding, reactor assembly/ disassembly, modifications and minor maintenance on ERCW piping and valves, snubber inspections, S/G eddy current testing, modifications to remove MIN-K insulation and the Appendix R multiple spurious operation scenarios resolution project. For the selected tasks the inspectors reviewed established dose goals and discussed assumptions regarding the bases for the current estimates with responsible ALARA planners. The inspectors evaluated the incorporation of exposure reduction initiatives and operating experience, including historical post-job reviews, into RWP requirements. Since the packages were from a previous outage the planned dose could be compared to actuals, adjustments made to planned doses could be retrospectively reviewed along with the basis of those adjustments. The post job reviews were available. Where applicable, the inspectors discussed changes to established estimates with ALARA planners and evaluated them against work scope changes or unanticipated elevated dose rates. With the ongoing transfer canal work the inspectors were able to review day to day dose tracking.

Source Term Reduction and Control The inspectors reviewed the collective exposure three-year rolling average from 2010 - 2012 and reviewed historical collective exposure trends. The inspectors evaluated historical dose rate trends and compared them to current data. Ongoing source term reduction initiatives such as ultrasonic fuel cleaning, zinc injection, resin overlays, elevated primary pH (7.3) were reviewed and discussed with HP staff. The licensees planned usage of X-ray fluorescent measurement to control the residual cobalt source terms being created by valve maintenance was discussed as well.

Radiation Worker Performance The inspectors had difficulty in finding radiologically significant work to observe due to the unit being on line with minimal ongoing repair work. The only meaningful observations were made of the personnel working in the transfer canal. While the inspectors were not able to observe any ALARA job briefs, they were able to observe job site work briefings. Radiation worker performance was also evaluated as part of IP 71124.01. While observing job tasks, the inspectors evaluated the use of remote technologies to reduce dose including teledosimetry and remote visual monitoring.

Problem Identification and Resolution The inspectors reviewed and discussed selected CAP documents associated with ALARA program implementation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedure NPG-SPP 03.1, Corrective Action Program, Rev. 6. The inspectors also evaluated the scope and frequency of the licensees self-assessment program and reviewed recent assessment results.

ALARA program activities were evaluated against the requirements of FSAR Section 12, TS Sections 5.7 (Programs and Procedures) and 5.11(High Radiation Areas), 10 CFR Part 20, and approved licensee procedures. Records reviewed are listed in Sections 2RS1 and 2RS2 of the report Attachment.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

Engineering Controls The inspectors reviewed the use of temporary and permanent engineering controls to mitigate airborne radioactivity during transfer canal repair activities.

The inspectors observed the use of negative pressure units (NPU)s and a temporary containment (tent) and reviewed NPU testing records. Use of containment purge to reduce airborne levels in general areas was reviewed. The inspectors evaluated the effectiveness of continuous air monitors and air samplers placed in work area breathing zones to provide indication of increasing airborne levels.

Respiratory Protection Equipment The inspectors reviewed the use of respiratory protection devices to limit the intake of radioactive material. This included review of devices used for routine tasks and devices stored for use in emergency situations. The inspectors reviewed ALARA evaluations for the use of respiratory protection devices during transfer canal repair. Selected Self-Contained Breathing Apparatus (SCBA) units and negative pressure respirators (NPR)s staged for routine and emergency use in the Main Control Room and other locations were inspected for material condition, SCBA bottle air pressure, number of units, and number of spare masks and air bottles available. The inspectors reviewed maintenance records for selected SCBA units for the past two years and evaluated SCBA and NPR compliance with National Institute for Occupational Safety and Health certification requirements. The inspectors also reviewed records of air quality testing for supplied-air devices and SCBA bottles.

The inspectors observed issuing respiratory protection devices to workers, including verification of training and medical qualifications. The inspectors discussed training for various types of respiratory protection devices with HP staff, observed fit testing a worker for an NPR and SCBA mask, observed use of powered air-purifying respirators (PAPR) for transfer canal work, and interviewed radworkers and control room operators on use of the devices including SCBA bottle change-out and use of corrective lens inserts. Respirator qualification records and medical fitness records were reviewed for several Main Control Room operators and emergency responder personnel in the Maintenance and HP departments. In addition, qualifications for individuals responsible for testing and repairing SCBA vital components were evaluated through review of training records.

Problem Identification and Resolution PERs associated with airborne radioactivity mitigation and respiratory protection were reviewed and assessed. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure NPG-SPP 03.1, Corrective Action Program, Rev. 6. Documents reviewed are listed in the Attachment.

Licensee activities associated with the use of engineering controls and respiratory protection equipment were reviewed against 10 CFR Part 20; UFSAR Chapter 12; the guidance in Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection; and applicable licensee procedures. Documents reviewed are listed in the Attachment.

The inspectors completed 1 sample as required by IP 71124.03 (sample size of 1).

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

a. Inspection Scope

External Dosimetry The inspectors reviewed the licensees National Voluntary Accreditation Program (NVLAP) certification data for accreditation for 2011-2012 and 2012-2013 for Ionizing Radiation Dosimetry. The inspectors reviewed program procedures for processing active personnel dosimeters (ED)s and onsite storage of Optically Stimulated Luminescent Dosimeters (OSLD)s. Comparisons between ED and OSLD results, including correction factors, were discussed in detail. The inspectors also reviewed dosimetry occurrence reports regarding alarming dosimeters.

Internal Dosimetry Inspectors reviewed and discussed the in vivo bioassay program with the licensee. Inspectors reviewed procedures that addressed methods for determining internal or external contamination, releasing contaminated individuals, the assignment of dose, and the frequency of measurements depending on the nuclides.

Inspectors reviewed and evaluated Whole Body Counter (WBC) sensitivity, count time and libraries. The inspectors discussed assessment and disposition of unexpected dosimetry results to include workers reporting for work after working abroad having received documentable uptakes. The inspectors evaluated the licensees program for in vitro monitoring, however there were no dose assessments using this method to review.

There were no internal dose assessments for internal exposure greater than 10 millirem committed effective dose equivalent to review.

Special Dosimetric Situations The inspectors reviewed records for three currently declared pregnant workers (DPW)s and discussed guidance for monitoring and instructing DPWs. Inspectors reviewed the licensees practices for monitoring external dose in areas of expected dose rate gradients, including the use of multi-badging and extremity dosimetry. The inspectors evaluated the licensees neutron dosimetry program including instrumentation which was evaluated under procedure 71124.05.

Problem Identification and Resolution The inspectors reviewed and discussed licensee CAP documents associated with occupational dose assessment. Inspectors evaluated the licensees ability to identify and resolve the identified issues in accordance with procedure NPG-SPP 03.1, Corrective Action Program, Rev. 6. The inspectors also discussed the scope of the licensees internal audit program and reviewed recent assessment results.

HP program occupational dose assessment activities were evaluated against the requirements of FSAR Section 12; TS Section 5.7; 10 CFR Parts 19 and 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

a. Inspection Scope

Radiation Monitoring Instrumentation: During tours of the auxiliary building, SFP areas, and RCA exit point, the inspectors observed installed radiation detection equipment including the following instrument types: area radiation monitors (ARM), continuous air monitors (CAM), liquid and gaseous effluent monitors, personnel contamination monitors (PCM), small article monitors (SAM), and portal monitors (PM). The inspectors observed the physical location of the components, noted the material condition, and compared sensitivity ranges with UFSAR requirements.

In addition to equipment walk-downs, the inspectors observed source checks and alarm setpoint testing of various portable and fixed detection instruments, including ion chambers, telepoles, PCM, SAM, and PM. For the portable instruments, the inspectors observed the use of a high-range calibrator and discussed periodic output value testing with a radiation protection technician. The inspectors reviewed the last two calibration records and evaluated alarm setpoint values for selected ARM, PCM, PM, SAM, effluent monitors, laboratory counting systems, and WBC systems. This included a sampling of instruments used for post-accident monitoring such as containment high-range ARMs, and effluent monitor high-range noble gas and iodine channels. Radioactive sources used to calibrate selected ARMs and effluent monitors were evaluated for traceability to national standards. Calibration stickers on portable survey instruments and air samplers were noted during inspection of storage areas for ready-to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to determine if calibration and check sources are representative of the plant source term. The inspectors also reviewed countroom quality assurance records for gamma ray spectroscopy equipment.

Effectiveness and reliability of selected radiation detection instruments were reviewed against details documented in the following: 10 CFR Part 20; NUREG-0737, Clarification of TMI Action Plan Requirements; TS Section 3.3; UFSAR Chapters 11 and 12; and applicable licensee procedures. Documents reviewed are listed in the Attachment.

Problem Identification and Resolution: The inspectors reviewed and discussed selected Corrective Action Program (CAP) documents associated with radiological instrumentation. The reviewed items included PERs, self-assessment, and quality assurance audit documents. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure NPG-SPP 03.1, Corrective Action Program, Rev. 6. Documents reviewed are listed in the Attachment.

The inspectors completed 1 sample as required by IP 71124.05 (sample size of 1).

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the five PIs listed below. To verify the accuracy of the PI data reported during the periods listed, PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 6, were used to verify the basis in reporting for each data element.

  • Mitigating System Performance Index (MSPI) - High pressure injection system
  • MSPI - Cooling water
  • MSPI - Emergency AC power Occupational Radiation Safety Cornerstone The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from June 2012 through July 2013. For the assessment period, the inspectors reviewed ED alarm logs and selected CRs related to controls for exposure significant areas. The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data. Documents reviewed are listed in the Attachment.

Public Radiation Safety Cornerstone The inspectors reviewed the Radiological Control Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone from June 2012 through July 201.

For the assessment period, the inspectors reviewed cumulative and projected doses to the public contained in liquid and gaseous release permits and PERs related to Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual issues.

The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data. Documents reviewed are listed in the Attachment.

The inspectors completed two of the required samples specified in Inspection Procedure (IP) 71151.

b. Findings

No findings were identified.

4OA2 Identification & Resolution of Problems

.1 Review of Items Entered into the CAP

As required by Inspection Procedure (IP) 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished by reviewing daily PER summary reports and periodically attending daily PER review meetings. Documents reviewed are listed in the Attachment.

.2 Annual Sample: Corrective Actions Associated with NCV 05000390/2011004-02,

Failure to Fully Implement Corrective Actions for a Motor Boat Necessary for Flood Mode Preparation

a. Inspection Scope

The inspectors reviewed the CAP, PER 417920, and the actions completed for NCV 05000390/2011004-02, Failure to Fully Implement Corrective Actions for a Motor Boat Necessary for Flood Mode Preparation. Documents reviewed are listed in the

.

b. Findings and Observations

No findings were identified.

4OA3 Event Follow-up

Unit 1 Reactor Trip - June 28, 2013

a. Inspection Scope

The inspectors reviewed control room activities following the reactor trip of June 28, 2013 as well as immediate trouble shooting activities and root cause investigation.

b. Findings and Observations

Introduction:

The inspectors monitored the licensee performing a trouble shooting plan following the reactor trip. The licensee attributed the cause of the reactor trip to a loose phase A connection on a digital fault recorder. The inspectors continued to monitor the root cause development following reactor startup to determine the validity of the cause and review the associated LER for closure.

Description:

On June 28, 2013, an A-phase high impedance ground fault occurred on the Roane 500kV transmission line approximately 22 miles from Watts Bar Nuclear Unit 1. Concurrently, the licensee experienced a reactor trip due to the actuation of the 1A Main Bank Transformer Feeder Differential Relay 187TF. The 500kV transmission line fault was caused by a tree that fell onto the A phase of the transmission line. The tree was cut by a local land owner. Operations personnel stabilized the plant using the Auxiliary Feedwater (AFW) System and the main steam dump valves. The secondary-side steam generator (SG) atmospheric relief valves, SG power operated relief valves (PORVs) and SG safety valves were not challenged during the transient. The Reactor Coolant System (RCS) responded to the initial plant transient as expected without actuating Pressurizer PORVs or initiating Safety Injection signals.

Per design, a differential relay, such as the 187TF relay, should not trip due to an event occurring outside of the relays zone of protection because the input amperage subtracts from the output amperage equaling zero so no amperage is available to trip the relay.

Specifically, the 187TF relay zone of protection covers the bus network between the two main generator output breakers and the 1A main bank transformer, which is within the plants switch yard, whereas the fault was 22 miles from the plant site. Initially, the licensee tried to verify that the differential circuits current transformers (CTs) were electrically balanced over their range of operation by injecting increasing levels of test amperage into the circuit. The CTs measure the amperage of the 500KV power and feed that measurement to the 187TF relay so verifying that the input CT amperage properly subtracts from the output CT amperage would validate the CTs characteristics.

Because the technicians injecting the amperage did not disable the 86 relay, all of the remaining circuit breakers connected to the X bus opened. The 86 relay detects if any breaker on the bus fails to open when called upon. The 86 relay operation had no significant effects on the shutdown of the plant, but because the technicians did not secure the circuit properly, TVA management decided to stop the verification of the CTs.

Alternatively, the technicians tried to verify circuit connections by physically moving the wiring. While handling an A phase wire connected to one of the Digital Fault Recorders, a technician stated that he noticed about 1/8 inch of movement and heard a click. The click is the locking mechanism that ensures the connection remains secure; however the design of the connector ensures electrical connection over one inch of movement.

Before the 86 relay tripped the remaining bus relays, the amperage injection had passed approximately 2 amps through this portion of the circuit, which would have detected a loose connection.

Inspectors observed that TVA stopped assessing other electrically significant reasons that could have tripped the 187TF relay such as CT imbalances. The root cause team determined that the click heard by the technician was the cause of the relay trip even though subsequent bench testing could not support it. In response to inspector questions, the licensee hired a 3rd party consultant, which also discounted the connector as an obvious cause of the 187TF relay trip. Both the inspectors and the 3rd party consultant believe that neither of the licensees troubleshooting techniques nor their root cause analysis has adequately addressed the cause of the 187TF relay trip, which can continue to challenge the reactor protection system on subsequent high impedance ground faults outside of the plant.

The licensee plans to disable the 187TF relay during the next shutdown for refueling, in the spring of 2014, in order to measure the current sensed by the relay as the main generator load is decreased for shut down. This item is identified as unresolved item (URI) 050000390/2013004-01, Contribution of Potential Current Transformer Imbalance to Reactor Trip.

4OA5 Other Activities

None

4OA6 Meetings, including Exit

On August 16, 2013, the inspectors discussed the results of the radiation safety inspection with Mr. Tim Cleary, Watts Bar Site Vice President, and other responsible staff. The inspectors noted that proprietary and sensitive information was reviewed during the course of the inspection and it would not be included in the documented report.

On October 2, 2013, the resident inspectors presented the quarterly inspection results to members of the licensee staff. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

4OA7 Licensee-Identified Violations

None ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

Y. Aboulfaida, Systems Engineering
R. Bankes, Chemistry/Environmental Manager
L. Belvin, QA Manager
G. Boerschig, Plant Manager
M. Casner, Site Engineering Director
T. Cleary, Site Vice President
S. Connors, Operations Manager
T. Detchemende, Emergency Preparedness Manager
K. Dietrich, Engineering Programs Manager
R. Dittmer, Operations Superintendent
D. Gronek, Plant Manager
D. Guinn, Licensing Manager
W. Hooks, Radiation Protection Manager
B. Hunt, Operations Support Superintendent
D. Jacques, Security Manager
T. Morgan, Licensing Engineer
D. Murphy, Maintenance Manager
W. Prevatt, Work Control Manager
R. Stroud, Site Licensing
M. Tuck, Radiation Protection

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

050000390/2013004-01 URI Contribution of Potential Current Transformer Imbalance to Reactor Trip (Section 4OA3)

Closed

05000390/2012005-03 URI Engineering Justification for Modifications to Non-Conforming Baskets (Section 1R18.2)

LIST OF DOCUMENTS REVIEWED

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