IR 05000390/2014005

IR 05000390/2014005 & 501; on 10/01/2014 - 12/31/2014; Watts Bar, Unit 1; Problem Identification and Resolution; Maintenance of Emergency Preparedness

ML15040A425
Person / Time
Site:	Watts Bar
Issue date:	02/09/2015
From:	Mark King Reactor Projects Region 2 Branch 6
To:	James Shea Tennessee Valley Authority
References
EA-14-197, IR 2014005, IR 2014501 IR 2014005, IR 2014501
Download: ML15040A425 (35)

Text

UNITED STATES ebruary 9, 2015

SUBJECT:

WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000390/2014005 and 05000390/2014501

Dear Mr. Shea:

On December 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Unit 1. On January 29, 2015, the NRC inspectors discussed the results of this inspection with Mr. Connors and other members of the Watts Bar staff. Inspectors documented the results of this inspection in the enclosed inspection report.

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

These findings involved violations of NRC requirements. Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance (Green)

in this report. The NRC is treating these violations as non-cited violations (NCV) 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, D.C. 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC resident inspector at the Watts Bar Nuclear Plant.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 II, and the NRC resident inspector at the Watts Bar Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 System (PARS)

component of 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/

Michael King, Chief Reactor Projects Branch 6 Division of Reactor Projects Docket No.: 50-390 License No.: NPF-90

Enclosure:

NRC Inspection Report 05000390/2014005, 05000390/2014501 w/Attachment: Supplemental Information

REGION II==

Docket No.: 50-390 License No.: NPF-90 Report No.: 05000390/2014005 05000390/2014501 Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Unit 1 Location: Spring City, TN 37381 Dates: October 1 through December 31, 2014 Inspectors: J. Nadel, Senior Resident Inspector R. Monk, Senior Resident Inspector J. Hamman, Resident Inspector R. Baldwin, Senior Operation Engineer A. Goldau, Operations Engineer S. Sanchez, Senior Emergency Preparedness Inspector M. Speck, Senior Emergency Preparedness Inspector C. Fontana, Emergency Preparedness Inspector Approved by: Michael King, Chief Reactor Projects Branch 6 Division of Reactor Projects Enclosure

SUMMARY

IR 05000390/2014-005 & 501; 10/01/2014 - 12/31/2014; Watts Bar, Unit 1; Problem

Identification and Resolution; Maintenance of Emergency Preparedness.

The report covered a three-month period of inspection by the resident inspectors and announced inspections by regional inspectors. One Green NRC-identified and one self-revealing findings were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," (SDP) dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 5.

Cornerstone: Mitigating Systems

Green: A self-revealing non-cited violation (NCV) of 10 CFR 50.54(q)(2) was identified for the licensees failure to maintain 1-RM-90-404 A/B radiation monitors as required by their Radiological Emergency Plan and NRC-approved Emergency Action Level scheme. The issue was placed in the licensees corrective action program and the radiation monitors are being replaced under a design change on an expedited schedule.

The licensees failure to identify the extended loss of the Unit 1 Watts Bar Nuclear Power Plants 1-RM-90-404 A&B CVE Radiation Monitors was not compliant with their approved emergency plan and was a failure to comply with 10 CFR 50.54(q)(2). This issue was more than minor because it was associated with the Facilities and Equipment attribute of the Reactor Safety - Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency in that equipment relied upon to prompt decision-makers to declare emergencies was not available. The finding was determined to have very low safety significance (Green) because alternative Emergency Action Levels within the radiological effluent initiating condition, facility design as documented in their Updated Final Safety Analysis Report, and the licensees declaration processes, were such that an accurate and timely declaration would have been made. This finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution component. The TVA failed to thoroughly evaluate the impact of the long term unavailability of the 1-RM-90-404A/B radiation monitors on the emergency plan. Specifically, since these radiation monitors are used to make site area and general emergency declarations, their continued long term unavailability presented a potentially safety significant concern. The licensee failed to identify this concern as an Emergency Preparedness issue and failed to take timely corrective actions to restore the failed radiation monitors (P.2). (Section 1EP5)

Green: An NRC-identified Green NCV of 10 CFR Part 50 Appendix B Criterion VII was identified for the licensees failure to assure that purchased chillers for the main control room (MCR) and shutdown board room (SDBR) conformed to the procurement documents. Specifically, the equipment qualification documentation provided for the MCR and SDBR chillers did not provide sufficient evidence to reasonably conclude the equipment would be able to perform its active safety function of heat control before, during, and after the analyzed safe shutdown earthquake.

The inspectors determined that the licensees failure to meet procurement specification SL M-0024-0 for qualification of MCR and SDBR chillers was a performance deficiency.

The cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was more than minor because, if left uncorrected, the condition had the potential to lead to a more significant safety concern.

The main control room chillers were required to assure habitability was maintained for mitigation and control of analyzed accidents. Because no further equipment qualification activities were planned, the capability of the main control room and shutdown board room cooling equipment to withstand an analyzed earthquake would continue to be indeterminate after installation and placement into service. The NRC concluded that the finding was of very low safety significance (Green) because the chillers had not been installed. The inspectors determined that the finding was directly related to the cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3). (Section 4OA2)

Licensee-Identified Violations

One licensee-identified violation of very low safety significance was reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violation and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

The unit started the reporting period at 100 percent rated thermal power (RTP) and remained there through the end of the reporting period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors reviewed licensee actions taken in preparation for low temperature weather conditions to limit the risk of freeze-related initiating events and to adequately protect mitigating systems from its effects. The inspectors reviewed licensee procedure 1-PI-OPS-1-FP, Freeze Protection, including associated checklist 1, Freeze Protection.

Inspectors also walked down selected components associated with the intake pumping station to evaluate implementation of plant freeze protection, including the material condition of insulation, heat trace elements, and temporary heated enclosures.

Corrective actions for items identified in relevant problem evaluation reports (PERs) and work orders (WOs) were assessed for effectiveness and timeliness. This activity constituted one Adverse Weather inspection sample.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors conducted the 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 (OOS). This also included that redundant trains were returned to service properly. The inspectors reviewed the functional system descriptions, the 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. This activity constituted three Equipment Alignment Partial Walkdown inspection samples.

• Partial walkdown of the 1A-A charging pump while the 1B-B pump was OOS for planned maintenance

• Partial walkdown of the 1A-A component cooling pump while the 1B-B pump was OOS for motor replacement

• Partial walkdown of the 1A-A emergency diesel generator (EDG) while the 1B-B EDG was OOS for surveillance testing

b. Findings

No findings were identified.

.2 Complete System Walkdown

a. Inspection Scope

The inspectors conducted a detailed walkdown/review of the alignment and condition of the auxiliary (flood mode) charging system 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/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. This activity constituted one Complete System Walkdown inspection sample.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope

The inspectors conducted tours of the 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; and 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 10 Fire Protection walkdown inspection samples.

• Cable spreading room

• 480 V reactor (RX) motor-operated valve (MOV) board room 1A

• 480 V RX MOV board room 1B

• 480 V RX MOV board room 2A

• 480 V RX MOV board room 2B

• Vital battery room I, II, III, IV, V (counts as five samples)

b. Findings

No findings were identified.

1R06 Flood Protection Measures

.1 Auxiliary Building

The inspectors reviewed internal flood protection measures for the auxiliary building flood protection features. The features were examined to verify that they were installed and maintained consistent with the plant design basis. The inspectors also reviewed the licensees flooding study calculation for determining maximum flood level in all building rooms for a limiting medium energy line break in the component cooling water system.

Inspectors also reviewed licensed and non-licensed operator response procedures to ensure operator actions could be accomplished. The inspectors confirmed that flood mitigation features such as drains, sumps, curbs and door seals were not degraded in such a manner as to adversely impact the conclusions of the study. This inspection constituted one Internal Flood Protection inspection sample.

.2 Cables in Underground Manholes

a. Inspection Scope

Inspectors directly observed, as listed below, the underground bunkers/manholes subject to flooding that contained cables whose failure could disable risk-significant equipment. Specific attributes evaluated were: the cables were not submerged in water, the cables and/or splices appeared intact and the material condition of cable support structures was acceptable, and dewatering devices (sump pump) operation and level alarm circuits were set appropriately to ensure that the cables would not be submerged or were in an environment for which they were qualified. Where dewatering devices were not installed, the inspectors ensured that drainage was provided and was functioning properly. This inspection constituted one Underground Manhole Internal flooding inspection sample.

• Manhole 9-B

• Manhole 15

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification and Performance

.1 Licensed Operator Requalification

a. Inspection Scope

On, November 18, 2014, the inspectors observed the simulator exam evaluation per scenario 3-OT-SRE-1010, 1B-B main feedwater pump trip and runback with a subsequent steam generator tube rupture. The plant conditions led to a Notification of Unusual Event and Alert classification. 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 assessed the licensees ability to administer testing and assess the performance of their licensed operators. The inspectors attended the post-examination critique performed by the licensee evaluators and verified that licensee-identified issues were comparable to issues identified by the inspector. The inspectors reviewed simulator physical fidelity (i.e., the degree of similarity between the simulator and the reference plant control room, such as physical location of panels, equipment, instruments, controls, labels, and related form and function). This activity constituted one Observation of Requalification Activity inspection sample.

b. Findings

No findings were identified.

.2 Control Room Observations

a. Inspection Scope

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 and other 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 This activity constituted one Control Room Observation inspection sample.

b. Findings

No findings were identified.

.3 Biennial Review of Licensed Operator Requalification

a. Inspection Scope

The inspectors reviewed the facility operating history and associated documents in preparation for this inspection. During the week of November 17, 2014, the inspectors reviewed documentation, interviewed licensee personnel, and observed the administration of operating tests associated with the licensees operator requalification program. Each of the activities performed by the inspectors was done to assess the effectiveness of the facility licensee in implementing requalification requirements identified in 10 CFR Part 55, Operators Licenses. The evaluations were also performed to determine if the licensee effectively implemented operator requalification guidelines established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and Inspection Procedure 71111.11, Licensed Operator Requalification Program. The inspectors also evaluated the licensees simulation facility for adequacy for use in operator licensing examinations using ANSI/ANS-3.5-1985, American National Standard for Nuclear Power Plant Simulators for use in Operator Training and Examination. The inspectors observed four crews during the performance of the operating tests. Documentation reviewed included written examinations, Job Performance Measures (JPMs), simulator scenarios, licensee procedures, on-shift records, simulator modification request records, simulator performance test records, operator feedback records, licensed operator qualification records, remediation plans, watchstanding records, and medical records. The records were inspected using the criteria listed in Inspection Procedure 71111.11. Documents reviewed during the inspection are documented in the List of Documents Reviewed.

This activity constituted one inspection sample.

b. Findings

No findings were identified.

.4 Annual Review of Licensee Requalification Examination Results:

a. Inspection Scope

On December 19, 2014, the licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), Requalification Requirements, of the NRCs Operators Licenses. The inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, Licensed Operator Requalification Program. These results were compared to the thresholds established in Section 3.02, Requalification Examination Results, of IP 71111.11. This activity constituted one inspection sample.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the performance-based problem 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) characterizing reliability issues for performance monitoring; 5) tracking unavailability for performance monitoring; 6) balancing reliability and unavailability; 7) trending key parameters for condition monitoring; 8) system classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); 9) appropriateness of performance criteria in accordance with 10 CFR 50.65(a)(2); and 10) appropriateness and adequacy of 10 CFR 50.65 (a)(1) goals, monitoring and corrective actions. This activity constituted two Maintenance Effectiveness inspection samples.

• PER 930382 - A train motor generator set room chiller thermostat control panel failure

• Review of evaluation of transferring System 31, SDBR ventilation, function 31K from a(1) to a(2)

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors evaluated, as appropriate, for the 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 activity constituted three Maintenance Risk Assessment inspection samples.

• Risk assessment for work week 1005 with a tornado watch issued for Rhea County

• Risk assessment for work week 1102 with a motor-driven auxiliary feedwater (MDAFW) pump and containment spray (CS) pump surveillance and crane work in the switchyard

• Risk Assessment for work week 1117 with essential raw cooling water (ERCW)maintenance, diesel generator (DG) 1A-A electric board damper emergent failure, and emergent trip risk from Reactor Protection System

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the 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; and 4) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation (LCO) and the risk significance in accordance with the SDP. The inspectors verified that the operability evaluations were performed in accordance with NPG-SPP-22.300, Corrective Action Program and OPDP-8, Operbility Dermination Process and Limiting Conditions for Operation Tracking. This activity constituted eight Operability Evaluation inspection samples.

• Past operabilitiy evaluation (POE) for PER 927995, Auxiliary Building Passive Sump drain path blocked

• PDO for PER 934675, RHR suction valve failure during recirculation

• PDO for PER 964936 NRC identified degraded Vimasco fire barrier in cable spreading room

• PDO for SR 949304 (PER 949565) steam generator No. 1 Inlet Check Valve Temperature Indication found out of "As Found" tolerance.

• Prompt determination of operability (PDO) for PER 751252, Boric Acid Residue from Leak on instrument line tubing below 1-ISIV-72-215C

• PER 951897 for a quality class 3 breaker found installed in a quality class 1 application

• Functional evaluation (FE) for PER 951617 for multiple HPFP supply valves to manual hose stations found mispositioned

• FE for PER 952103 for the discovery of a sinkhole and tailwater sediment at the Boone Dam.

b. Findings

No findings were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors reviewed the 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. This activity constituted five Post Maintenance Testing inspection samples.

• WO 115693175, 1-SI-70-901-A, Component cooling water pump 1A-A quarterly performance test following routine planned maintenance

• WO 116258951, Digital filter replacement following failure in 1-R-6 RPS protection rack

• WO 114458253 Thermal barrier booster pump Presray barrier test

• WO 116113745, replacement of the temperature control valve for the A MCR chiller

• WO 116258857, Component cooling water pump 1B-B shaft replacement

1R22 Surveillance Testing

a. Inspection Scope

The inspectors witnessed the 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. This activity constituted seven Surveillance Testing inspection samples: two in-service, four routine tests, and one RCS leakage sample.

In-Service Test:

• WO 115696657, 1-SI-62-901-B, Centrifugal charging pump quarterly performance test

• WO 116182994, 1-FCV-70-153, RHR heat exchanger 1B-B outlet valve, testing (PER 937184)

Other Surveillances:

• WO 115696943, 0-SI-82-12-A, 2A-A EDG Monthly Surveillance test

• WO 115697011, 1-SI-72-901-A, Containment spray pump 1A-A quarterly performance test

• WO 115717055, Monthly diesel generator start and load test DG 2B-B

• WO 115811974 Verify/adjust power supplies and filter inspection protection set II RCS Leakage:

• WO 1144711808 (Rack portion) and WO 114472549 (Transmitter portion), 1-SI-77-1, 18 Month Channel Calibration Reactor Building Auxiliary Floor and Equipment Drain Pocket Sump Level, Loop 1-LPL-77-410 (counts as one sample)

b. Findings

No findings were identified.

Cornerstone: Emergency Preparedness

1EP2 Alert and Notification System Evaluation

a. Inspection Scope

The inspectors evaluated the adequacy of the licensees methods for testing the alert and notification system in accordance with NRC Inspection Procedure 71114, 02, Alert and Notification System (ANS) Testing. The applicable planning standard, 10 CFR Part 50.47(b)(5), and its related 10 CFR Part 50, Appendix E, Section IV.D requirements were used as reference criteria. The criteria contained in NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, were also used as a reference.

The inspectors reviewed various documents which are listed in the Attachment.

Inspectors interviewed personnel involved with siren system maintenance and observed the condition of a sample of siren installations. This inspection activity satisfied one inspection sample for the alert and notification system on a biennial basis.

b. Findings

No findings were identified.

1EP3 Emergency Preparedness Organization Staffing and Augmentation System

a. Inspection Scope

The inspectors reviewed the licensees Emergency Response Organization (ERO)augmentation staffing requirements and process for notifying the ERO to ensure the readiness of key staff for responding to an event and timely facility activation. The qualification records of key position ERO personnel were reviewed to ensure all ERO qualifications were current. A sample of problems identified from augmentation drills or system tests performed since the last inspection was reviewed to assess the effectiveness of corrective actions.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, 03, Emergency Preparedness Organization Staffing and Augmentation System. The applicable planning standard, 10 CFR 50.47(b)(2), and its related 10 CFR 50, Appendix E requirements were used as reference criteria.

This inspection activity satisfied one inspection sample for the ERO staffing and augmentation system on a biennial basis.

b. Findings

No findings were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

Since the last NRC inspection of this program area, Revisions 102 and 103 were made to the Radiological Emergency Plan. The licensee determined that, in accordance with 10 CFR 50.54(q), the plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The inspectors reviewed the changes and sampled implementing procedure changes made between October 2013 and September 2014 to evaluate for potential reductions in the effectiveness of the plan. As this review was not documented in a safety evaluation report and does not constitute formal NRC approval of the changes, these changes remain subject to future NRC inspection in their entirety.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, 04, Emergency Action Level and Emergency Plan Changes. The applicable planning standards of 10 CFR 50.47(b), and its related requirements in 10 CFR 50, Appendix E, were used as reference criteria.

This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.

b. Findings

No findings were identified.

1EP5 Maintenance of Emergency Preparedness

a. Inspection Scope

The inspectors reviewed the corrective actions identified through the Emergency Preparedness (EP) program to determine the significance of the issues, the completeness and effectiveness of corrective actions, and to determine if issues were recurring. The licensees post-event action reports, self-assessments, and audits were reviewed to assess the licensees ability to be self-critical, thus avoiding complacency and degradation of their EP program. Inspectors reviewed the licensees 10 CFR 50.54(q) change process, personnel training, and selected screenings and evaluations to assess adequacy. The inspectors toured facilities and reviewed equipment and facility maintenance records to assess licensees adequacy in maintaining them. The inspectors evaluated the capabilities of selected radiation monitoring instrumentation to adequately support Emergency Action Level (EAL) declarations.

The inspection was conducted in accordance with NRC Inspection Procedure 71114.05, Maintenance of Emergency Preparedness. The applicable planning standards, related 10 CFR 50, Appendix E requirements, and 10 CFR 50.54(q) and

(t) were used as reference criteria.

This inspection activity satisfied one inspection sample for the maintenance of EP on a biennial basis.

b. Findings

Introduction:

A self-revealing Green NCV of 10 CFR 50.54(q)(2) was identified for the failure to follow and maintain the effectiveness of the licensees emergency plan.

Specifically, from April 2010 until the present, Unit 1 Condenser Vacuum Exhaust (CVE) Mid/High Range Monitors (1-RM-90-404 A&B RMs) were unavailable to perform their EP function to provide EAL inputs to classify emergency action levels, RG1.1, General Emergency (GE), and RS1.1, Site Area Emergency (SAE). This condition could challenge plant operations when determining emergency classification for gaseous effluent release during a steam generator tube leak or rupture.

Description:

The Watts Bar Radiological Emergency Plan (REP) includes EAL initiating conditions RS1.1 and RG1.1 for radiological effluent releases. The EALs list several radiation monitors and associated threshold values for the 1-RM-90-404 A&B Condenser Vacuum Exhaust (CVE) Radiation Monitors. These monitors along with 1-RM-90-119 (CVE Low-Range monitor) continuously monitor the main condenser exhaust effluent release path and function to monitor for steam generator leakage. The low-range monitor, main steam line radiation monitors, steam generator blowdown effluent radiation monitors, and abnormal operating/emergency operating procedures as well as effluent path sample points continued to be available. In 2006, 1-RM-90-404 began to experience frequent failures and required significant efforts to keep in service. In 2009, the licensee determined that the monitor should be replaced due to unreliability and obsolescence and a formal design change was considered. In April 2010 and July 2011, the design change was delayed due to lack of funding even though the monitor was still not functioning. In March 2012, corporate EP implemented actions to track and resolve EP equipment issues in the form of a department procedure then eventually a higher level fleet corporate procedure. In more fully implementing the corporate procedure in November 2013, the licensees failure to maintain the 1-RM-90-404 radiation monitors since 2010 was self-revealed. The licensee placed this failure in their CAP, formally codified the compensatory actions in a standing order, raised the priority in implementing the design change process, and performed an Apparent Cause Evaluation. The CVE Mid/High Range Radiation Monitoring (1-RM-90-404 A&B) instruments remain unavailable however the design change is currently being implemented.

Analysis:

The inspectors determined that the failure to identify the extended loss of the Unit 1 Watts Bar Nuclear Power Plants 1-RM-90-404 A&B CVE Radiation Monitors was not compliant with their approved emergency plan and was a failure to comply with 10 CFR 50.54(q)(2), was within the licensees ability to foresee and correct, and therefore constituted a performance deficiency. This finding was determined to be more than minor because it was associated with the Facilities and Equipment attribute of the Reactor Safety - Emergency Preparedness Cornerstone. This finding adversely affected the cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency in that equipment relied upon to prompt decision-makers to declare emergencies was not available.

This finding was evaluated in accordance with Inspection Manual Chapter (IMC) 0609, Appendix B, Emergency Preparedness Significance Determination Process. The issue was determined to be a Failure to Comply and was evaluated using Section 5.0 and 2, Failure to Comply Significance Logic, With regards to EALs, RG1.1 and RS1.1, the inspectors considered mitigating factors, and determined that alternative EALs within the radiological effluent initiating condition, facility design as documented in their UFSAR, and the licensees declaration processes, were such that an accurate and timely declaration would have been made. Therefore, the inspectors determined that the Risk-Significant Planning Standard (RSPS) function was neither lost nor degraded and determined to be of low safety significance (Green).

A cross-cutting aspect, Evaluation (P.2), was identified within the Problem Identification and Resolution component. The TVA failed to thoroughly evaluate the impact of the long term unavailability of the 1-RM-90-404A/B radiation monitors on the emergency plan.

Specifically, since these radiation monitors are used to make site area and general emergency declarations, their continued long term unavailability presented a potentially safety significant concern. The licensee failed to identify this concern as an EP issue and failed to take timely corrective actions to restore the failed radiation monitors.

Enforcement:

Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part and the planning standard of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4), states, A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. 1-RM-90-404A/B Site Area and General Emergency threshold values are included in the licensees NRC-approved EAL scheme for radiological effluent. Contrary to the above, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, since April 2010, the licensee failed to restore regulatory compliance by not maintaining 1-RM-90-404 A&B Mid/High Range Radiation Monitors, which are used to classify EALs, RG1.1, General Emergency, and RS1.1, Site Area Emergency. The 1-RM90-404 A&B Mid/High Range Radiation Monitoring Detectors are still unavailable however, replacement is in progress on an expedited schedule and means to identify alternate EALs within the radiological effluent initiating condition are available such that an accurate and timely declaration can be made therefore the condition does not present an immediate safety concern. Because this finding was of low safety significance (Green) and was entered into the licensees CAP as PER 807406, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy and is identified as (NCV)05000390/2014501-01, Failure to Maintain Effectiveness of Emergency Plan.

4OA1 Performance Indicator Verification

a. Inspection Scope

The inspectors sampled licensee submittals relative to the performance indicators (PIs)listed below for the period July 1, 2013, through June 30, 2014. To verify the accuracy of the PI data reported during that period, PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, were used to confirm the reporting basis for each data element.

Emergency Preparedness Cornerstone

• ERO Drill/Exercise Performance

• ERO Drill Participation

• Alert and Notification System Reliability For the specified review period, the inspectors examined data reported to the NRC, procedural guidance for reporting PI information, and records used by the licensee to identify potential PI occurrences. The inspectors verified the accuracy of the PI for ERO drill and exercise performance through review of a sample of drill and event records.

The inspectors reviewed selected training records to verify the accuracy of the PI for ERO drill participation for personnel assigned to key positions in the ERO. The inspectors verified the accuracy of the PI for alert and notification system reliability through review of a sample of the licensees records of periodic system tests. The inspectors also interviewed the licensee personnel who were responsible for collecting and evaluating the PI data. Licensee procedures, records, and other documents reviewed within this inspection area are listed in the Attachment. This inspection satisfied three inspection samples for PI verification on an annual basis.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Review of items entered into the Corrective Action Program:

a. Inspection Scope

As required by Inspection Procedure 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 licenseesCAP. This review was accomplished by reviewing daily PER and service request (SR) reports, and periodically attending Corrective Action Review Board and PER Screening Committee meetings.

b. Findings

No findings were identified.

.2 Semiannual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, the inspectors performed a review of the licensees CAP and other associated programs and documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also included licensee trending efforts and licensee human performance results. The inspectors review nominally considered the six-month period of July through December 2014, although some examples expanded beyond those dates when the scope of the trend warranted. Inspectors reviewed licensee trend reports for the period in order to determine the existence of any adverse trends that the licensee may not have previously identified. The inspectors review also included the licensees integrated trend reports. The inspectors verified that adverse or negative trends identified in the licensees PERs, periodic reports and trending efforts were entered into the CAP. This inspection satisfied one inspection sample for Semiannual Trend Review.

b. Findings and Observations

No findings were identified. In general, the licensee had identified trends and appropriately addressed them in their CAP. The inspectors observed that the licensee had performed a detailed review. The licensee routinely reviewed cause codes and utilized key words and system links to identify potential trends in their data. The majority of licensee identified trends were identified by the quality assurance organization over this time period. Inspectors had observed and commented on the same trends that the licensee was identifying, particularly the negative trends in fire protection program performance and emergency drill controller performance. The inspectors compared the licensee process results with the results of the inspectors daily screening. No new adverse trends were identified this period that had not already been identified by the licensee

.3 Annual Sample: Review of Operator Workarounds

a. Inspection Scope

The inspectors reviewed the operator workaround program to verify that workarounds were identified at an appropriate threshold, were entered into the CAP, and that corrective actions were proposed or implemented. Specifically, the inspectors reviewed the licensees workaround list and repair schedules, conducted tours, and interviewed operators about required compensatory actions. Additionally, the inspectors looked for undocumented workarounds, reviewed appropriate system health documents, and reviewed PERs related to items on the workaround list. This activity constituted one Operator Workaround inspection sample

b. Findings and Observations

No findings were identified

.4 Annual Sample: Review of ERCW Strainer 1A History of Leakage (PER 699727)

a. Inspection Scope

The inspectors performed an in-depth review of the licensees corrective actions associated with PER 699727, ERCW Strainer 1A History of Leakage. The inspectors assessed, as appropriate, the licensees problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensees corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors also evaluated the PER against the requirements of the of the licensees CAP as specified in procedure NPG-SPP-22.300, Corrective Action Program, and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the CAP for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.

b. Findings and Observations

No findings of significance were identified. The subject PER identified that a long history of packing leakage, some in excess of 5gpm, had been occurring on the 1A ERCW strainer. Past corrective action to repack the strainer had not been successful and the subject PER identified that strainer shaft should be replaced to correct the condition.

The inspectors reviewed the history of past SRs and PERs on the strainer leakage and discovered that the leakage had been documented in the CAP at least since 2001.

Inspectors noted that many of the older SRs written on the leakage did not result in PERs being written and did not have operability assessments performed as required. As a result, the multiple SRs written about excessive packing leakage between 2001 and 2010 were all closed to WOs that were subsequently canceled. This contributed to the long timeline before the damaged shaft was identified and corrective actions were created in PER 699727. Inspectors noted that the planned shaft replacement corrective actions from PER 699727 have not been completed due to boundary isolation challenges with the ERCW system and are currently planned for the next Unit 1 outage as a result. The inspectors concluded that the above issues were not more than minor because the packing leakage does not impact the ability of the strainer or the ERCW system to perform its safety function. This issue was documented in the licensees CAP program as PER 956629.

.5 Annual Sample: Corrective Actions Associated with PER 630349, Issues with Seismic

Test Data and Reporting

a. Inspection Scope

The inspectors interviewed responsible licensee engineering and project management, and reviewed documentation related to the disposition of PER 630349, Issues with Seismic Test Data and Reporting, issued October 25, 2012. The PER initiator stated that various revisions of seismic test data and reporting provided to TVA contained conflicting information and did not provide clear objective evidence of satisfactory test completion. The inspectors evaluated whether the identified condition was properly characterized as to significance, and whether sufficient corrective actions were identified and implemented to correct any associated conditions adverse to quality.

b. Findings and Observations

Introduction:

An NRC-identiified Green NCV of 10 CFR Part 50 Appendix B Criterion VII was identified for the licensees failure to assure that purchased chillers for the MCR and SDBR conformed to the procurement documents. Specifically, the equipment qualification documentation provided for the MCR and SDBR chillers did not provide sufficient evidence to reasonably conclude the equipment would be able to perform its active safety function of heat control before, during, and after the analyzed safe shutdown earthquake.

Description:

The inspectors reviewed the disposition provided for PER 630349, which documented a concern that records for equipment qualification of replacement chillers for the main control room and shutdown board room did not show that testing had met specified requirements. The applicable qualification requirements were outlined in procurement specification SL M-0024-0, Shutdown Board Room Water Chiller A-A and B-B. Section 211 of the specification required the complete package chiller to be seismically designed and constructed to operate during and after a safe shutdown earthquake in accordance with WB DC-40-31.2. Section 301.D required the chiller control system to maintain chilled water temperature at the desired setpoint, specified in Section 201 as 42 degrees Fahrenheit. The inspectors determined the PER had been closed by stating that sufficient documentation was provided in revision 2 of equipment qualification report N1673-SEISMI-QUAL-1. However, the inspectors review of the equipment qualification report and interviews with the responsible licensee engineer and project managers determined the following:

• Records of initial factory testing conducted prior to shipment for seismic testing did not provide information that was sufficient to show the chiller unit under test successfully controlled temperature within a specified range.

• Records of subsequent seismic shake testing documented that the evaporator water circuit was dry and did not maintain a flow of chilled water during the test evolutions.

As a result, the capability of the chiller to control evaporator chilled water outlet temperature was not tested.

• Records of a special test conducted by the manufacturer several months after completion of seismic shake tests disclosed that the chiller would not start without disabling a required element of the trip circuitry. Once started, the chiller failed to control evaporator outlet temperature within the specified range.

The inspectors determined that a safety function to provide heat control was identified in Safety Classification Evaluation MR1259475, Replacement Shutdown Board Room Chillers - System 031. In addition, Watts Bar General Design Criteria Document WB-DC-40-31.2, Seismic/Structural Qualification of Seismic Category I Electrical and Mechanical Equipment, Requirement 3.1, stated the program for seismic qualification of Seismic Category I equipment at WBN shall conservatively demonstrate that no loss of function (as defined by the TVA engineering discipline responsible for the equipment functional design) shall result either before, during, and/or after the occurrence of the postulated seismic events and other Watts Bar design basis loading conditions for which functionality must be assured.

Analysis:

The inspectors determined that the licensees failure to meet procurement specification SL M-0024-0 for qualification of MCR and SDBR chillers was a performance deficiency. Specifically, the equipment qualification documentation provided for the MCR and SDBR chillers did not provide sufficient evidence to reasonably conclude the equipment would be able to perform its active safety function of heat control before, during, and after the analyzed safe shutdown earthquake.

The performance deficiency was more than minor because if left uncorrected, the condition had the potential to lead to a more significant safety concern. The MCR chillers were required to assure habitability was maintained for mitigation and control of analyzed accidents. Because no further equipment qualification activities were planned, the capability of the MCR and SDBR cooling equipment to withstand an analyzed earthquake would continue to be indeterminate after installation and placement into service.

The NRC concluded that the finding was of very low safety significance (Green) because the chillers had not been installed. The failures to meet specified requirements for qualification of the MCR and SDBR chillers had been previously self-identified and evaluated under PER 630349 without correcting the deficiencies. Accordingly, the inspectors determined that the finding was directly related to the cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).

Enforcement:

10 CFR Part 50 Appendix B, Criterion VII, states, in part, that measures be established to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures shall include provisions, as appropriate, for objective evidence of quality furnished by the contractor or subcontractor.

Contrary to the above, on June 3, 2013, measures implemented by the licensee for acceptance of the final equipment qualification report for the safety-related MCR and SDBR chillers did not assure that the purchased chillers conformed to the procurement documents. Specifically, the qualification package furnished by the contracted supplier and approved by the licensee did not contain objective evidence of quality, in that the documentation failed to show the equipment would be able to perform its active safety function of heat control before, during, and after the analyzed safe shutdown earthquake. This does not constitute an immediate safety concern because the MCR chillers were not installed for use. Because this finding was of low safety significance (Green) and was entered into the licensees CAP as PER 813679, this violation is being treated as a NCV, consistent with the NRC Enforcement Policy and is identified as (NCV)05000390/2014005-01, Failure to Assure Qualification Testing Met Specified Requirements.

4OA6 Meetings, including Exit

On January 29, 2014, the resident inspectors presented the quarterly inspection results to Mr. Connors and other members of the licensee staff. The inspectors confirmed that none of the potential report inputs discussed were considered proprietary.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation.

10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to the above, on November 10, 2014, the licensee failed to accomplish procedure TI-64, Breaching Hazard Barriers, Revision 8 in accordance with its instructions. On November 10, 2014, during unit 2 component cooling water system flow balance testing, personnel used a wooden wedge to block open door A029, 1-B-B Charging Pump Room Door, in the auxiliary building. This was done to facilitate tubing connections needed for the testing.

TVA procedure TI-64, Breaching Hazard Barriers, Revision 8, Appendix A, Barrier Breach Matrix explicitly states that door A029 cannot be breached while the unit is in mode 1. The licensee identified this condition and took immediate actions to secure the door. The 1B-B charging pump was declared inoperable for the approximately 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period of time the door was breached. Using IMC 0609, Appendix A, Exhibit 2 (Mitigating Systems); this finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of function of at least a single train of equipment for greater than its technical specification allowed outage time.

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

G. Arent, Licensing Manager

R. Bankes, Chemistry/Environmental Manager

L. Belvin, QA Manager

M. Bottorff, Operations Superintendent

M. Casner, Site Engineering Director

S. Connors, Plant Manager

K. Dietrich, Manager Engineering Programs

T. Detchemendy, Emergency Preparedness Manager

S. Fisher, Security Manager

L. Freeman, Project Manager

H. Hammati-Arass, Manager Projects

D. Heidrich, Mechanical Engineer

W. Hooks, Radiation Protection Manager

J. James, Maintenance Manager

T. Morgan, Licensing Engineer

J. ODell, Site Licensing Supervisor

A. Pirkle, Engineering Programs

J. Reidy, Operations Manager

D. Shutt, Licensing

R. Stroud, Site Licensing

M. Thaggart, Work Control Manager

K. Walsh, Site Vice President

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000390/2014501-01 (NCV) Failure to Maintain Effectiveness of Emergency Plan (Section 1EP5)

05000390/2014005-01 (NCV) Failure to Assure Qualification Testing Met Specified Requirements (Section 4OA2.5)