IR 05000390/2009002
| ML091140240 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 04/24/2009 |
| From: | Heather Gepford Reactor Projects Region 2 Branch 6 |
| To: | Swafford P Tennessee Valley Authority |
| References | |
| IR-09-002 | |
| Download: ML091140240 (38) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931
April 24, 2009
Mr. Preston D. Swafford Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000390/2009002 AND 05000391/2009002, AND ANNUAL ASSESSMENT MEETING
SUMMARY
Dear Mr. Swafford:
On March 31, 2009, the United States Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Units 1 and 2. The enclosed integrated inspection report documents the inspection results which were discussed on April 7 and April 15, 2009, with Mr. G. Boerschig and Mr. M. Skaggs, respectively, and other members of your 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.
Based upon the results of this inspection, no findings of significance were identified. One licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of its very low safety significance and because it was entered into your corrective action program, the NRC is treating this licensee-identified violation as a non-cited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy.
If you contest the NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Watts Bar facility.
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/
Heather J. Gepford, Acting Chief
Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos. 50-390, 50-391 License No. NPF-90 and Construction Permit No. CPPR-92
Enclosure: NRC Inspection Report 05000390/2009002, 05000391/2009002 w/Attachments cc w/encl: (See page 3)
_________________________
X SUNSI REVIEW COMPLETE OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RI:DRS SIGNATURE RM MP RL DMP PH HG JH NAME RMonk MPribish RLewis DMasPenaranda PHiggins HGepford JHamman DATE 4/21/2009 4/21/2009 4/24/2009 4/22/2009 4/24/2009 4/24/2009 4/24/2009 E-MAIL COPY?
YES NO YES NO YES NO YES NO YES NO YES NO YES NO
TVA cc w/encl:
Gordon P. Arent New Generation Licensing Manager Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution
Masoud Bajestani, Vice President Watts Bar Nuclear Plant Unit 2 Tennessee Valley Authority Electronic Mail Distribution
Ashok S. Bhatnagar, Sr. Vice President Nuclear Generation Development and Construction Tennessee Valley Authority Electronic Mail Distribution
Michael K. Brandon, Manager Licensing and Industry Affairs Watts Bar Nuclear Plant Electronic Mail Distribution
Preston D. Swafford Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority Electronic Mail Distribution
William R. Campbell, Sr. Vice President Fleet Engineering Tennessee Valley Authority Electronic Mail Distribution
Thomas Coutu, Vice President Nuclear Support Tennessee Valley Authority Electronic Mail Distribution
General Counsel Tennessee Valley Authority Electronic Mail Distribution
Ludwig E. Thibault, General Manager Nuclear Oversight & Assistance Tennessee Valley Authority 3R Lookout Place 1101 Market Steet Chattanooga, TN 37402-2801
Gregory A. Boerschig, Plant Manager Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution
Larry E. Nicholson, General Manager Performance Improvement Tennessee Valley Authority Electronic Mail Distribution
Michael A. Purcell Senior Licensing Manager Nuclear Power Group Tennessee Valley Authority Electronic Mail Distribution
Michael J. Lorek, Vice President Nuclear Engineering & Projects Tennessee Valley Authority Electronic Mail Distribution
Michael D. Skaggs, Site Vice President Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution
Fredrick C. Mashburn, Acting Manager Corp. Nuclear Licensing and Industry Affairs Tennessee Valley Authority Electronic Mail Distribution
Senior Resident Inspector Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission 1260 Nuclear Plant Road Spring City, TN 37381-2000
County Executive 375 Church Street Suite 215 Dayton, TN 37321
County Mayor P.O. Box 156 Decatur, TN 37322
Lawrence E. Nanney, Director Division of Radiological Health TN Dept. of Environment & Conservation Electronic Mail Distribution
James H. Bassham, Director Tennessee Emergency Mgmt. Agency Electronic Mail Distribution
Ann Harris 341 Swing Loop Rockwood, TN 37854
Letter to Preston D. Swafford from Heather J. Gepford dated April 24, 2009
SUBJECT:
WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000390/2009002 AND 05000391/2009002
Distribution w/encl:
C. Evans, RII EICS L. Slack, RII EICS OE Mail RIDSNRRDIRS PUBLIC RidsNrrPMWattsBar1 Resource RidsNrrPMWattsBar2 Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-390, 50-391
License Nos:
NPF-90 and Construction Permit CPPR-92
Report Nos:
05000390/2009002, 05000391/2009002
Licensee:
Tennessee Valley Authority (TVA)
Facility:
Watts Bar Nuclear Plant, Units 1 and 2
Location:
Spring City, TN 37381
Dates:
January 1, 2009 - March 31, 2009
Inspectors:
R. Monk, Senior Resident Inspector M. Pribish, Resident Inspector J. Hamman, Reactor Inspector, RII (Section 1R17)
P. Higgins, Project Engineer, RII (Sections 1R04, 1R06 and 1R18)
R. Lewis, Senior Reactor Inspector, RII (Section 1R17 D. Mas-Penaranda, Reactor Inspector, RII (Section 1R17)
T. Fanelli, Trainee (Section 1R17)
Approved by:
Heather J. Gepford, Acting Chief Reactor Projects Branch 6 Division of Reactor Projects
SUMMARY OF FINDINGS
IR 05000390/2009-002, 05000391/2009-002; 01/01/2009 - 03/31/2009; Watts Bar, Units 1 & 2;
Integrated Inspection Report.
The report covered a three-month period of routine inspection by the resident inspectors and announced inspections by a regional project engineer, a senior reactor inspector, and a reactor inspector. No findings of significance were identified. 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.
A violation of very low safety significance that was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective action tracking number are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period with the exception of a planned power reduction to 46 percent RTP on March 20, 2009, to plug leaking tubes in the main condenser. The unit returned to full power on March 24, 2009 and remained there for the remainder of the inspection period.
Restart of construction on Unit 2 began in December of 2007. Information on Watts Bar Unit 2 reactivation can be found at http://www.nrc.gov/reactors/plant-specific-items/watts-bar.html
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
==1R01 Adverse Weather Protection
a. Inspection Scope
==
The inspectors reviewed the licensees preparation for and response to an actual freezing condition on January 15, 2009. The inspectors verified performance and reviewed the data associated with temperature monitoring of the refueling water storage tank (RWST), which is required per licensee procedure 1-PI-OPS-1-FP for outside air temperature less than 25 degrees F. In addition, the inspectors performed a walkdown of the RWST freeze protection enclosures to verify the adequacy of construction and the operation of the installed temporary lighting and temperature monitoring system.
b. Findings
No findings of significance were identified.
==1R04 Equipment Alignment
==
.1 Partial 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. 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.
- B-train containment spray (CS) system during A-train CS component outage
- A-train auxiliary building gas treatment system (ABGTS) while the B-train ABGTS out of service for maintenance
- Vital Battery II while aligned to the 6-S battery charger during emergent work on the normal battery charger
b. Findings
No findings of significance were identified.
.2 Semiannual Complete System Walkdown
a. Inspection Scope
The inspectors conducted one detailed walkdown/review of the alignment and condition of the auxiliary feedwater system (AFW) 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 to ensure that the licensee was identifying equipment alignment problems and that they were properly addressed for resolution. Further, various operating experience documents and reports were reviewed to identify if this experience was utilized and addressed by the licensee. This inspection sample was completed using the guidance listed in Operating Experience Smart Sample FY 2009-02. Documents reviewed are listed in Attachment 1 to this report.
b. Findings
No findings of significance were identified.
Introduction:
The inspectors identified an unresolved item (URI) involving compliance of the AFW System with General Design Criterion 2. This URI will remain unresolved pending additional information from the licensee to determine if a violation of regulatory requirements occurred.
Description:
On March 9-11, 2009, inspectors performed a routine inspection of the AFW system in accordance with Inspection Procedure 711111.04 "Equipment Alignment". This inspection also used the insights gained in Operating Experience Smart Sample (OpESS)
"Negative Trend and Recurring Events Involving Feedwater Systems". During their review of operating experience, the inspectors noted that a failure of an AFW pumps had occurred at the Callaway Plant in 2002 due to debris generated from a degraded diaphragm in the condensate storage tank (CST).
The inspectors examined system diagrams and confirmed that at Watts Bar there were no screens or other devices which would prevent debris from the CSTs from impacting the safety related auxiliary feed water pumps. The inspectors noted that the CSTs were not safety-related and were not protected from earthquakes or tornadoes. The inspectors also determined that the CSTs were lined with an epoxy phenolic coating which was routinely repaired during plant outages. However, if this coating became dislodged in sufficient quantities during normal operations or if an earthquake or other natural phenomena would cause CST debris, the AFW pumps could be adversely impacted. The UFSAR identified the AFW system as a safety-related system which is protected from the effects of natural phenomena. The UFSAR also stated that the plant complies with the requirements of General Design Criterion (GDC) 2, in that, structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes and floods. Pending additional information from the licensee involving potential for CST debris and its impact on the AFW system and compliance with GDC 2, this item is identified as URI 050000390/2009002-01, Auxiliary Feedwater System Compliance with General Design Criterion 2.
==1R05 Fire Protection
==
.1 Fire Protection - Tours
a. Inspection Scope
The inspectors conducted tours of the eight areas important to reactor safety, listed below, to verify the licensees implementation of fire protection requirements as described in the Fire Protection Program, Standard Programs and Processes (SPP)-10.0, Control of Fire Protection Impairments, SPP-10.10, Control of Transient Combustibles, and SPP-10.11, 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.
- Vital DC Boardroom I
- Vital DC Boardroom II
- Vital DC Boardroom III
- Vital DC Boardroom IV
- A 6.9 KV shutdown board (SDBR)
- B 6.9 KV SDBR
- Motor-driven AFW pumps/CCS (component cooling system) pumps
b. Findings
No findings of significance were identified.
==1R06 Flood Protection Measures
a. Inspection Scope
==
The inspectors reviewed internal flood protection measures for the emergency diesel generator building. Flood protection features were examined to verify that they were installed and maintained consistent with the plant design basis. The inspectors also reviewed the licensee flooding study calculation for determining maximum flood level in all building rooms for piping failures in both the essential raw cooling water (ERCW) system and the fire protection system and confirmed that flood mitigation features such as drains and curbs were not degraded in such a manner as to adversely impact the conclusions of the study. Documents reviewed are listed in Attachment 1 to this report.
b. Findings
No findings of significance were identified.
==1R11 Licensed Operator Requalification
a. Inspection Scope
==
On January 22, 2009, the inspectors observed the simulator evaluations for scenario 3-OT-SRTECA-3.2-13, Steam Generator Tube Rupture and Loss of Coolant Accident (LOCA).
The plant conditions led to a site area emergency level classification.
The inspectors specifically evaluated the following attributes related to the 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 inspectors.
b. Findings
No findings of significance were identified.
==1R12 Maintenance Effectiveness
a. Inspection Scope
==
The inspectors reviewed the two performance-based problems listed below. The focus of the reviews was 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 SPP-6.6, 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).
- Main turbine/generator a(1) to a(2) classification
- Thermal barrier booster pump a(1) performance improvement plan
b. Findings
No findings of significance were identified.
==1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
==
The inspectors evaluated, as appropriate for the four 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); SPP-7.0, Work Control and Outage Management; SPP-7.1, Work Control Process; and TI-124, Equipment to Plant Risk Matrix.
- Probabilistic risk assessment (PRA) evaluation WBN1-09-004 for workweek 09-702
- PRA evaluation WBN1-09-005 R2 for workweek 09-703
- PRA evaluation WBN1-09-007 for workweek 09-704
- Maintenance risk associated with replacement of 1A safety injection pump suction relief valve, 1-RFV-63-511
b. Findings
No findings of significance 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: and
- (4) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation (LCOs) and the risk significance in accordance with the NRC's Significance Determination Process. The inspectors verified that the operability evaluations were performed in accordance with SPP-3.1, Corrective Action Program.
- Problem Evaluation Report (PER) 159448, Failure of normal main control room pressure to be maintained
- PER 157755, Glycol chiller performance degradation
- PER 159025, B-EBR chiller oil cooler TCV leakage
- PER 155509, Tags in containment
- PER 162448, GE magne-blast breaker spring charging motor
b. Findings
No findings of significance were identified.
==1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
a.
==
Inspection Scope
The inspectors reviewed selected samples of evaluations to confirm that the licensee had appropriately considered the conditions under which changes to the facility, Updated Final Safety Analysis Report (UFSAR), or procedures may be made, and tests conducted, without prior NRC approval. The inspectors reviewed evaluations for seven changes and additional information, such as drawings, calculations, supporting analyses, the UFSAR, and Technical Specifications (TS) to confirm that the licensee had appropriately concluded that the changes could be accomplished without obtaining a license amendment. The seven evaluations reviewed are listed in Attachment 1 to this report.
The inspectors reviewed samples of changes for which the licensee had determined that evaluations were not required, to confirm that the licensees conclusions to screen out these changes were correct and consistent with 10CFR50.59. The 18 screened out changes reviewed are listed in Attachment 1 to this report.
The inspectors evaluated engineering design change packages for eleven material and design based modifications to evaluate the modifications for adverse effects on system availability, reliability, and functional capability. The twelve modifications and the associated attributes reviewed are as follows:
DCN 51934, Replacement of Internals of WBN-1-FCV-62-93 and -89, Rev. A (Initiating Events)
- Materials/Replacement Components
- Flowpath
- Process Medium
DCN 52309, Modifications to Station Control and Service Air Systems, Rev. A (Mitigating Systems)
- Materials/Replacement Components
- Pressure Boundary
- Flow Path
- Process Medium
DCN 52467, ABSCE (auxiliary building secondary containment envelope) Boundary Station Service Air System Mod to Support U2 Construction, Rev. A (Mitigating Systems)
- Flowpath
- Licensing Basis
CGD MR1101611, Compression Lug, #2 AWG Nickel Plated Copper for System 068, 11/20/07, (Mitigating Systems)
- Materials/Replacement Components
- Licensing Basis
CGD MR1138806, Soldered Split-Ferrule Cable Connector, 2/25/08 (Mitigating Systems)
- Materials/Replacement Components
- Licensing Basis
DCN 52300, Operate Diesel Fans Periodically, 1/11/08 (Mitigating Systems)
- Energy Needs
- Timing
- Control Signals
- Equipment Protection
- Operations
- Licensing Basis
DCN 52606, Move Safety Related Unit 1 / Unit 2 Interface Boundary Points for 125V DC Vital Power Battery Boards I through IV, 7/15/08 (Mitigating Systems)
- Equipment Protection
- Operations
- Licensing Basis
AA 52257, SI Pump Vent Lines, 11/29/07 (Mitigating Systems)
- Energy Needs
- Materials
- Operations
- Flowpath
- Pressure Boundary
- Ventilation Boundary
- Structural
- Failure Modes
- Licensing Basis DCN 52266, Change breaker instantaneous setting from LO to 2, Rev. A (Mitigating System)
- Energy Needs
- Process Medium
- Equipment Protection
- Post Modification Test
DCN 52625, Replace temperature switch for TDAFWP room DC vent fan, Rev. A (Mitigating Systems)
- Material Replacement Components
- Control Signals
- Energy Needs
- Post Modification Test
EDC 52884, Revise GSDS #3113 to encompass Boric Acid Blender Primary Water Supply Flow, Rev. A (Mitigating Systems)
- Material Replacement Components
- Control Signals
- Operations
- Post Modification Test
DCN 52388, Revise and modify software code in the Plant Integrated Computer System, Rev. A (Mitigating Systems)
- Timing
- Control Signal
- Operations
- Heat Removal
Documents reviewed included procedures, engineering calculations, modification design and implementation packages, work orders, site drawings, corrective action documents, applicable sections of the living UFSAR, supporting analyses, Technical Specifications, and design basis information. The inspectors additionally reviewed test documentation to ensure adequacy in scope and conclusion. The inspectors review was also intended to verify that all details were incorporated in licensing and design basis documents and associated plant procedures.
The inspectors also reviewed selected PERs and the licensees recent self-assessment associated with modifications and screening/evaluation issues to confirm that problems were identified at an appropriate threshold, were entered into the corrective action process, and appropriate corrective actions had been initiated and tracked to completion.
Additionally, given the licensees pursuits to complete construction of Unit 2, the team made inquiry into the status of previous 10 CFR 50.59 Evaluations and Screenings which had based their conclusions upon the unfinished state of Unit 2s facilities.
It was confirmed that there exists some fractional percentage of previous screenings and evaluations which will need to be revisited in light of future plans and, in some cases, near-term activities.
The licensee briefed the team on how this was to be accomplished under construction interface procedures tied to licensing basis preservation and initiated a tracking PER (PER 162088) to ensure that there is a clear reference to this activity within their corrective action program.
b. Findings
.1 Acceptability of Plant Alterations Without NRC Submittal
Introduction:
A URI was identified following NRC review of a licensee decision not to provide a submittal in association with temporary alteration TACF 1-07-0002-065, implemented in March 2007 on the Emergency Gas Treatment System (EGTS).
Description:
The temporary alteration (TACF 1-07-0002-065) changed the system start logic such that both trains would operate in automatic upon receiving the start signal. In their review of the licensees 10CFR50.59 evaluation, the inspectors found that the licensee had made a determination that no license amendment was required, though the supporting paragraph indicated a license amendment was appropriate and was to be accomplished in association with the licensees corrective action program (CAP).
Upon further review, the inspectors determined that, as part of a 2005 functional evaluation, a single-failure vulnerability was determined to exist. That same evaluation also recognized that the dose consequence for the system, described in the UFSAR, was based on single fan operation. Given the calculation of record assumptions that all fan flow would be exhausted and the fact that both system fans could achieve flow via that path, the licensee determined that there was an increased beta dose consequence over that described in the UFSAR to Main Control Room operators.
NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, limits beta dose in the control room to 30 rem. Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59, Section 4.3.3, states that the criterion for a more than minimal increase in consequence is greater than 10% of the margin to the limit. For this specific case, based on the UFSAR beta dose value of 6.803 rem, an increase of greater than 2.320 rem [(30 rem - 6.803 rem)*0.1] would indicate a more than minimal increase in the consequence. Using the licensee's re-evaluated beta dose value of 9.757 rem, the consequence difference was determined to be a 2.954 rem increase in beta dose for main control room operators, which exceeds the NEI 96-07 criterion for a more than minimal increase. The licensee's determination of the need for a license amendment was carried as an action item in PER 91670.
Between 2005 and 2007 the licensee installed two temporary modifications to address the single-point failure vulnerability. The resulting configuration, at the time of inspection, did not resolve, nor compensate for the flow conditions which resulted in the licensees determination of the need for a submittal as of 2005. During the 2009 inspection, the inspectors found that no licensee amendment request had been submitted to the NRC to support changing the UFSAR while the calculations of record continued to indicate that one was warranted.
This issue was unresolved pending additional NRC review to assess the adequacy of the licensees actions in response to the functional evaluation and the adequacy of 10 CFR 50.59 evaluations associated with the related temporary EGTS configuration modifications.
This item is identified as URI 05000390/2009002-02: Acceptability of Plant Alterations Without NRC Submittal.
.2 Acceptability of Seismic Qualification of 120VAC Vital Instrumentation Board Heinemann
Circuit Breakers
Introduction:
A URI was identified related to the adequacy of the seismic qualification of the station 120VAC Vital Instrumentation Boards.
Description:
The licensee originally procured their 120VAC Vital Instrumentation Boards as a functional unit, dedicated and seismically qualified by a third-party qualifier. In the early 1990s, the licensee implemented a complete replacement of the Heinemann circuit breakers in the instrumentation boards with commercial grade breakers from the same manufacturer. A different third party was contracted to verify seismic qualifications for the replacement breakers. Based upon documentation provided for review, the inspectors were unable to determine that the licensee had provided either the appropriate seismic data as procurement requirements for the circuit breakers or sufficient information to allow determination of the amplification factors to support application/location specific testing to this new third party qualifier. Further, the factor that was used as a ground force multiplier did not appear to have sufficient basis so as to ensure that it bounded peak values that the breakers might experience.
Additionally, the inspectors noted that the mounting for the 120VAC Vital Instrumentation Boards was not consistent with the documented seismic testing breaker mounting. While both the licensees and the third party qualifiers standards cite compliance with IEEE Standard 344, the inspectors were concerned that the breaker test mounting did not dynamically simulate the plant-specific mounting, nor bound the full spectrum of panel response contributors (Paragraph 7.4 of IEEE 344, 1987 version).
These factors, taken together, caused the inspectors to question the suitability of the subcomponent qualification in that the tested response spectrum did not appear to fully bound the required response spectrum across the ground frequency range.
The licensee has additionally noted that the manufacture of the subject breakers has recently undergone substantial variation in form and fit. This issue is being tracked within the licensees CAP. The breakers are not currently installed in the subject panels, but the licensee was considering panel modifications to allow for the utilization of the new breakers.
This issue will remain open for further NRC review of the adequacy of the licensees seismic qualification of the 1990s installed replacement circuit breakers in the 120VAC Vital Instrumentation Boards. This item is identified as URI 05000390/2009002-03:
Acceptability of Seismic Qualification of 120VAC Vital Instrumentation Board Circuit Breakers.
==1R18 Plant Modifications
1. Temporary Plant Modifications
a. Inspection Scope
==
The inspectors reviewed the following temporary plant modifications against the requirements of SPP-9.5, Temporary Alterations, and SPP-9.4, 10 CFR 50.59 Evaluation of Changes, Test, and Experiments, and to verify that the modifications did not affect system operability or availability as described by the TS and UFSAR.
In addition, the inspectors determined whether:
- (1) the installation of the temporary modification was in accordance with the work package;
- (2) adequate configuration control was in place;
- (3) procedures and drawings were updated; and,
- (4) post-installation tests verified operability of the affected systems.
- WO 08-822382-000: Installation of a recorder which was directly attached to various terminal points inside a safety related motor control center to troubleshoot and monitor circuits which are related to the auxiliary oil pump associated with the 1A-A Centrifugal Charging Pump.
- WO 08-822382-001: Installation of a tee connection in the oil system of the 1A-A Centrifugal Charging Pump to allow a connection of a temporary pressure transmitter and recorder and installation of a recorder which was directly connected to an oil system pressure switch to troubleshoot and monitor performance of the auxiliary oil system.
b. Findings
No findings of significance were identified.
2. Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed one permanent plant modification to verify that design change installation controls were adequate; affected operational procedures and licensing documents were identified and revised accordingly; and that post-maintenance testing and equipment return to service was adequate. Documents reviewed are listed in Attachment 1 to this report.
- Design change notice (DCN) 52307-Installation of a blank off plate to isolate ERCW flow to one of the eight cooling coils associated with Lower Compartment Cooler (LCC) 1D-B.
b. Findings
No findings of significance were identified.
==1R19 Post-Maintenance Testing
a. Inspection Scope
==
The inspectors reviewed six 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 SPP-8.0, Testing Programs; SPP-6.3, Pre-/Post-Maintenance Testing; and SPP-7.1, Work Control Process.
- WO 09-810836-000 and WO 09-810836-001, Functional test of A and B train MDAFW (motor driven AFW) auto suction rollover from CST (condensate storage tank) to ERCW
- WO 07-811354-000, 1-FCV-67-298, Upper containment ventilation cooler D isolation valve inside MOVATS test
- WO 07-8231969-020, ABGTS drawdown test to confirm adequacy of the relocated ABSCE boundary per DCN 52283
- WO 09-812400, Diesel fire pump discharge relief valve pressure control setup
- WO 03-016475-016, Replace 30 RX relay on WBN-1-BKR-211-1718/11-A, Maintenance supply from 6.9kv Unit BD (board) 1B
b. Findings
No findings of significance 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; SPP-8.0, Testing Programs; SPP-8.2, Surveillance Test Program; and SPP-9.1, 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.
Routine Surveillance Tests:
- WO 09-811538-000, 0-SI-67-923-B, Essential raw cooling water pump G-B preservice test
- 1-SI-0-21, Excore QPTR (quadrant power tilt ratio) and axial flux difference
In-Service Test:
- WO 08-820260-000, 1-SI-3-902, Turbine driven auxiliary feedwater pump 1A-S quarterly performance test
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
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 Emergency Plan Implementing 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 99-02, Regulatory Assessment Performance Indicator Guideline.
- Anticipated transient without scram (ATWS) and loss of coolant accident (LOCA)lead to site area emergency and a general emergency
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verifications
a. Inspection Scope
The inspectors sampled licensee submittals for the four 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 5, were used to verify the basis in reporting for each data element.
Initiating Events Cornerstone PI
- Unplanned scrams per 7000 critical hours
- Unplanned scrams with complications
- Unplanned power changes per 7000 critical hours
The inspectors reviewed selected licensee event reports and portions of the operator logs from the period of January 1, 2008, to December 31, 2008, to verify that the licensee had accurately identified the number of scrams and unplanned power changes greater than 20 percent that occurred during the previous four quarters. The inspectors also reviewed the accuracy of the number of critical hours reported and the licensees basis for including or excluding each scram in the unplanned scrams with complications PI.
Barrier Integrity Cornerstone PI
- Reactor coolant system activity
The inspectors reviewed portions of the operator and chemistry logs from the period of January 1, 2008, to December 31, 2008, to verify that the licensee had accurately determined and reported the reactor coolant system maximum dose equivalent iodine-131 activity during the period reviewed.
b. Findings
No findings of significance were identified.
4OA2 Identification & Resolution of Problems
.1 Review of Items Entered into the Corrective Action Program
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 licensees CAP. This review was accomplished by reviewing daily PER summary reports and attending daily PER review meetings.
.2 Annual Sample: Corrective actions associated with non-cited violation (NCV)
05000390/2006004-03, Inappropriate Crediting of Operator Restoration Actions Causes Inadequate Risk Assessment
a. Inspection Scope
The inspectors reviewed the plan and implementation of corrective actions for NCV 05000390/2006004-03, which was documented in PER 108886. PER 108886 contained corrective actions to determine if training on licensee procedure TI-124, Equipment to Plant Risk Matrix, was needed and to implement the identified training. Additionally, the inspectors reviewed the effectiveness of the licensees processes and procedures for assessing and managing risk associated with maintenance activities in accordance with paragraph (a)(4) of 10 CFR 50.65, Requirements For Monitoring The Effectiveness Of Maintenance At Nuclear Power Plants.
b. Findings and Observations
No findings of significance were identified. However, the inspectors identified several observations which were discussed with the licensee.
Paragraph (a)(4) of 10 CFR 50.65 states, in part, that before performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Licensee procedure TI-124 is the licensees guiding procedure for assessing and managing risk in accordance with paragraph (a)(4) of 10 CFR 50.65. In accordance with TI-124, risk assessments may be performed using the ORAM-sentinel risk assessment computer program or the matrix attached to TI-124. TI-124 follows the guidelines of NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, and establishes risk thresholds based on incremental core damage probability (ICDP) and incremental large early release probability (ILERP). The risk categories are established as Green, Yellow, Orange and Red. Risk management plans and approval are required for any risk level other than Green.
The inspectors determined the procedures and processes in place were adequate to effectively implement the requirements of paragraph (a)(4) of 10 CFR 50.65. In addition, training was conducted in this area as described in PER 108886; however, over the past two years, the inspectors have noted a continuing lack of rigor with the implementation of TI-124 and (a)(4) requirements as noted by NRC observations which were documented by the licensee in the following PERs:
- PER 124269: Transition to Yellow risk was overlooked during the 1B RHR component outage. NRC Finding 05000390/2007003-01, Inadequate Risk Assessment for Work In Progress.
- PER 130342: The remaining portion of 0-SI-31-7-B, Control Room Ventilation Filter Test, was completed without being fully analyzed for risk prior to performance. This item was determined to be minor because the activity, when properly analyzed for risk, was determined to be a Green risk activity.
- PER 141982: Breaker swap risk review was not performed as stated. This item was determined to be minor because the activity, when properly analyzed for risk, was determined to be a Green risk activity.
- PER 161964: Risk assessment for work week 702 did not include WO 08-818630-004. When properly analyzed for risk, the activity was determined to be a Yellow risk condition. This item was determined to be minor because the activity was delayed and did not occur in work week 702.
- PER 164966: The point at which a risk condition is entered is not programmatically addressed and as a result, the licensee did not realize the plant conditions established had placed them into an analyzed Yellow condition. This item was determined to be minor because the activity was conservatively determined to be Yellow by the licensee, but was, in fact, a Green risk activity based on the ICDP.
- PER 165076: A weakness in the process to identify and perform all associated risk management actions was identified when one of the prescribed risk management actions was not taken. This item was determined to be minor because the activity was conservatively determined to be Yellow by the licensee, but was, in fact, a Green risk activity based on the ICDP.
4OA3 Event Followup
a. Inspection Scope
The inspectors reviewed the events surrounding the 2A emergency diesel generator (EDG)output breaker failure that occurred following monthly surveillance testing on January 30, 2009. Initial investigation revealed that the breakers spring charging motor had failed (PER 162448). The inspectors verified through direct observation and document review that actions taken by the licensee to determine the extent of condition were adequate to ensure continued operability of potentially affected breakers.
b. Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.
b. Findings
No findings of significance were identified.
.2 (Closed) Temporary Instruction (TI) 2515/176, EDG TS Surveillance Requirements
Regarding Endurance and Margin Testing
a. Inspection Scope
Inspection activities for TI 2515/176 were previously completed and documented in inspection report 05000390, 391/2008004, and this TI is considered closed at Watts Bar; however, TI 2515/176 will not expire until August 31, 2009. The information gathered while completing this temporary instruction was forwarded to the Office of Nuclear Reactor Regulation for review and evaluation.
b. Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
.1 Integrated Inspection Report Meeting Exit
The inspectors presented the inspection results to Mr. G. Boerschig and other members of licensee management at the conclusion of the inspection on April 7, 2009. A subsequent exit was conducted via conference call on April 15, 2009, with Mr. M. Skaggs and other members of licensee management. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Annual Assessment Meeting Summary
On April 14, 2009, Dr. Heather Gepford, Acting Branch Chief, Region II, met with Mr. Greg Boerschig, Plant Manager, and other members of the licensee staff to discuss the NRCs annual assessment of the Watts Bar Nuclear Plant's safety performance for the period of January 1 through December 31, 2008. The annual assessment results were previously provided to TVA via letter dated March 4, 2009.
On April 14, 2009, the NRCs Acting Chief of Reactor Projects Branch 6, and the Resident Inspectors, held a Category 3 meeting for members of the public and local officials. This Category 3 public meeting provided an open house public forum to fully engage the public in a discussion of the NRCs Reactor Oversight Process and annual assessment of the Watts Bar Nuclear Plant's safety performance for the period January 1 through December 31, 2008. The members of the public expressed no concerns about the operation of the Watts Bar Unit 1 facility. The presentation material used for discussions and the list of attendees are attached to this report.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
- 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, 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, procedures, or drawings. Contrary to this, on February 27, 2009, the lower containment particulate radiation monitor was returned to service with incorrect alarm setpoints which rendered the particulate radiation monitor inoperable. This was identified in the licensees CAP as PER 164765. This finding is of very low safety significance because other methods of reactor coolant system leak detection were available.
ATTACHMENTS:
1. Supplemental Information
2. Attendance List
3. Meeting Presentation
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- L. Belvin, Radiation Protection Manager
- G. Boerschig, Plant Manager
- M. Brandon, Licensing and Industry Affairs Manager
- D. Helms, Senior Engineer
- B. Hunt, Operations Superintendent
- G. Mauldin, Site Engineering Manager
- M. McFadden, Site Nuclear Assurance Manager
- M. Pope, Licensing Engineer
- C. Riedl, Licensing Supervisor
- A. Scales, Operations Manager
- M. Skaggs, Site Vice President
- D. Voeller, Maintenance and Modifications Manager
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
URI Auxiliary Feedwater System Compliance with General
Design Criterion 2 (Section 1R04).
URI Acceptability of Plant Alterations Without NRC
Submittal (Section 1R17)
URI Acceptability of Seismic Qualification of 120VAC Vital
Instrumentation Board Circuit Breakers
(Section 1R17.2)
Closed
2515/176
TI EDG TS Surveillance Requirements Regarding Endurance and Margin Testing (Section 4OA5.2)
Discussed
None