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| | issue date = 04/20/2007 | | | issue date = 04/20/2007 |
| | title = IR 05000390-07-006, on 02/12/2007 - 02/16/2007, 02/26/2007 - 03/9/2007; Watts Bar Nuclear Power Plant; Component Design Bases Inspection | | | title = IR 05000390-07-006, on 02/12/2007 - 02/16/2007, 02/26/2007 - 03/9/2007; Watts Bar Nuclear Power Plant; Component Design Bases Inspection |
| | author name = Cain L M | | | author name = Cain L |
| | author affiliation = NRC/RGN-II/DRS/EB1 | | | author affiliation = NRC/RGN-II/DRS/EB1 |
| | addressee name = Swafford P D | | | addressee name = Swafford P |
| | addressee affiliation = Tennessee Valley Authority | | | addressee affiliation = Tennessee Valley Authority |
| | docket = 05000390 | | | docket = 05000390 |
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| =Text= | | =Text= |
| {{#Wiki_filter: | | {{#Wiki_filter:April 20, 2007Tennessee Valley AuthorityATTN:Mr. Preston D. Swafford, ActingChief Nuclear Officer and Executive Vice President6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801SUBJECT:WATTS BAR NUCLEAR PLANT- NRC COMPONENT DESIGN BASISINSPECTION REPORT 05000390/2007006 |
| [[Issue date::April 20, 2007]]
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| Tennessee Valley AuthorityATTN:Mr. Preston D. Swafford, ActingChief Nuclear Officer and Executive Vice President6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
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| SUBJECT: WATTS BAR NUCLEAR PLANT- NRC COMPONENT DESIGN BASISINSPECTION REPORT 05000390/2007006
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| ==Dear Mr. Swafford:== | | ==Dear Mr. Swafford:== |
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| TVA2In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, 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 NRC's 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). | | TVA2In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, 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 NRC's 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). |
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| Sincerely,/RA/ | | Sincerely, |
| | /RA/ |
| Loyd M. Cain, Acting ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos.:50-390License Nos.:NPF-90 and Construction Permit No. CPPR-92 | | Loyd M. Cain, Acting ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos.:50-390License Nos.:NPF-90 and Construction Permit No. CPPR-92 |
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| ===Enclosure:=== | | ===Enclosure:=== |
| NRC Inspection Report 05000390/2007006 | | NRC Inspection Report 05000390/2007006 w/Attachment: Supplemental Information |
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| ===w/Attachment:===
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| Supplemental Informationcc w/encl: (See Page 3) | |
| TVA3cc w/encl:Ashok S. Bhatnagar Senior Vice President Nuclear Operations Tennessee Valley Authority Electronic Mail DistributionLarry S. Bryant, Vice PresidentNuclear Engineering & Technical Services Tennessee Valley Authority Electronic Mail DistributionMichael D. SkaggsSite Vice President Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail DistributionPreston D. SwaffordSenior Vice President Nuclear Support Tennessee Valley Authority Electronic Mail DistributionGeneral CounselTennessee Valley Authority Electronic Mail DistributionJohn C. Fornicola, General ManagerNuclear Assurance Tennessee Valley Authority Electronic Mail DistributionBeth A. Wetzel, ManagerCorporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801Robert H. Bryan, Jr., General ManagerLicensing & Industry Affairs Tennessee Valley Authority Electronic Mail DistributionJames D. Smith, Acting ManagerLicensing and Industry Affairs Watts Bar Nuclear PlantTennessee Valley AuthorityElectronic Mail DistributionMichael J. Lorek, Plant ManagerWatts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail DistributionCounty ExecutiveRhea County Courthouse 375 Church Street, Suite 215 Dayton, TN 37321-1300County MayorP. O. Box 156 Decatur, TN 37322Lawrence E. Nanney, DirectorTN Dept. of Environment & Conservation Division of Radiological Health Electronic Mail DistributionAnn Harris341 Swing Loop Rockwood, TN 37854 James H. Bassham, DirectorTennessee Emergency Management Agency Electronic Mail Distribution
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| ___OFFICERII:DRSRII:DRSRII:DRSRII:DRScontractorcontractorRII: DRPSIGNATURE/RA//RA S.Rose for//RA//RA//RA S. Rose for//RA S.Rose for//RA/NAMESRoseWFowlerRBerrymanLCainHCampbellJLeivoMWidmannDATE04/17/200704/18/200704/19/200704/18/200704/19/200704/17/2007April , 2007 E-MAIL COPY? NO NO NO YES NO NO YES U.S. NUCLEAR REGULATORY COMMISSIONREGION IIDocket Nos.:50-390License Nos.:NPF-90 Report Nos.:05000390/2007006 Licensee:Tennessee Valley Authority Facility:Watts Bar Nuclear Plant, Unit 1 Location:Spring City, TN 37381 Dates:February 12 - March 9, 2007 Inspectors:S. Rose, Senior Reactor Inspector (Lead)R. Berryman, Senior Reactor Inspector W. Fowler, Reactor Inspector H. Campbell, Contractor J. Leivo, ContractorApproved by:Loyd M. Cain, Acting Chief, Engineering Branch 1 Division of Reactor Safety 2Enclosure
| | REGION IIDocket Nos.:50-390License Nos.:NPF-90 Report Nos.:05000390/2007006 Licensee:Tennessee Valley Authority Facility:Watts Bar Nuclear Plant, Unit 1 Location:Spring City, TN 37381 Dates:February 12 - March 9, 2007 Inspectors:S. Rose, Senior Reactor Inspector (Lead)R. Berryman, Senior Reactor Inspector W. Fowler, Reactor Inspector H. Campbell, Contractor J. Leivo, ContractorApproved by:Loyd M. Cain, Acting Chief, Engineering Branch 1 Division of Reactor Safety 2Enclosure |
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| =SUMMARY OF FINDINGS= | | =SUMMARY OF FINDINGS= |
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| ==DOCUMENTS REVIEWED== | | ==DOCUMENTS REVIEWED== |
| Updated Final Safety Analysis ReportUFSAR Section 9.4, Air Conditioning, Heating, Cooling, and Ventilation SystemsUFSAR Section 9.4.5.2.1, Diesel Generator Building | | Updated Final Safety Analysis ReportUFSAR Section 9.4, Air Conditioning, Heating, Cooling, and Ventilation SystemsUFSAR Section 9.4.5.2.1, Diesel Generator Building |
| : UFSAR Section 9.4.5.3, Auxiliary Building Engineered Safety Features (ESF) Equipment Coolers
| |
| : UFSAR Section 9.2.1, Essential Raw Cooling Water (ERCW)
| |
| : UFSAR Section 10.4.9, Auxiliary Feedwater System
| |
| : UFSAR Table 3.9-20, Relief Valves in Class 2 Auxiliary Systems
| |
| : UFSAR section 15.2.8, Loss of Normal FeedwaterCalculationsWBNOSG4136, Steady State DBE LOCA Temperatures for the Auxiliary Building, Rev. 14AB-030A-01, HVAC Air Balance, Rev. 4
| |
| : WBN-30-D053, DGB Ventilation, Static Pressure & Equipment Performance for Electrical Panel Fan, Rev. 7
| |
| : WBNOSG4136, Steady State DBE LOCA Temperatures for the Auxiliary Building, Rev. 14
| |
| : EPM-JTB-103189, Calculation Method for Analyzing HVAC Cooling Coil Performance, Rev. 2
| |
| : EPMGRB092992, Evaluation - 6900V Shutdown Board Room Chiller Under Varying Load and Condenser (ERCW) Water Temperature, Rev. 2
| |
| : WBNOSG4-071, RWST and Containment RHR Safety Limits, Analytical Limits and Setpoints, Rev. 17
| |
| : MDQ00106320060110, CCP, SIP, CSP, and RHR Pump NPSH Evaluation, Rev. 0
| |
| : EPM-GRB-092992, Determine 6.9 KV SDBR Chiller Conderser Safe Operating Pressure Range, Rev. 2
| |
| : EPMCES092689, Equipment Performance for the 6900 V Shutdown Board Room HVAC
| |
| : System, Rev. 8
| |
| : WBNOSG4136, Steady State DBE LOCA Temperatures for the Auxiliary Building, Rev. 14
| |
| : WBN-MEB-M-D-Q-0-031-00-0048, Cooling Load Analysis for Rooms Served by the Shutdown and 480V Board and Battery Room HVAC Systems, Rev. 0
| |
| : EPM-JFL-060395, ERCW Pump Lift Clearances, Rev. 0
| |
| : EPMPTC120594, Essential Raw Cooling Water System Pressure Drop Calculation, Rev. 6
| |
| : EPM-JFL-20285, ERCW System Flow Requirements, Rev. 8
| |
| : E31840612001, 6.9 kV Shutdown Board Essential Raw Cooling Water Pumps, Rev. 2
| |
| : E31840810005, 6.9 kV Shutdown Board Centrifugal Charging Pumps, Rev. 1
| |
| : EPMGDU041593, Brake Horsepower Analysis for Safety Related Components, Sections 1 - 9, Rev. 19
| |
| : WBNEEBEDQ1999010001, Auxiliary Power System Analysis, Rev. 38
| |
| : WBN-EEB-MS-TI03-0012, Diesel Generator Loading Analysis, Sections 1 - 9, Appendix B, Rev. 67
| |
| : WBN-EEB-MS-TI08-0008, 480V 1E Coordination Protection, Sections 1, 4, 5.1.2, 5.1.10.2,
| |
| : 6.1.9; Appendix 3, pp. 1, 14A; Appendix 8, p.7, Rev. 129
| |
| : WBN-EEB-MS-TI111-0062, 125 Vdc Diesel Generator Control Power System Evaluation, Rev. 21
| |
| : 3AttachmentWBNEEBMSTI120016, 120 Vac Vital Inverter Loading, Sections 1 - 9, Rev. 108WBPE2119202001, 6.9 kV Shutdown & Logic Boards Undervoltage Relays Requirement /
| |
| : Demonstrated Accuracy Calculation, Rev. 5
| |
| : WBPE2689301008, Permanent Hydrogen Mitigation System (PHMS) 120 Vac Power Supply Analysis, Rev. 0
| |
| : 1-FT-70-81A/81E, Demonstrated Accuracy Calculation for 1-FT-70-81A / 81E, Rev. 5Operating ProceduresE-1, Loss of Reactor or Secondary Coolant, Rev. 14ECA 0.0, Loss of Shutdown Power, Rev. 19
| |
| : ARI-125-B, Containment Hi-Hi Pressure Spray Actuate, Rev. 11
| |
| : FR-Z.1, High Containment Pressure Emergency Operating Instruction, Rev. 10
| |
| : I-SI-268-2-A, 18 Month Permanent Hydrogen Mitigation System Train A Operability Test, Rev. 4
| |
| : I-SI-268-2-B, 18 Month Permanent Hydrogen Mitigation System Train B Operability Test, Rev. 4
| |
| : AOI-10, Loss of Control Air, Rev. 38
| |
| : SOI-32.02, Auxiliary Air System, Rev. 18
| |
| : MI-17.021, Installation of Spool Pieces Between ERCW System and Component Cooling System, Rev. 6
| |
| : AOI-13, Loss of Essential Raw Cooling Water, Rev. 34
| |
| : AOI-40, Station Blackout, Rev. 10
| |
| : SOI-3.02, Auxiliary Feedwater System, Rev. 44
| |
| : I-SI-3-64, 18 Month Channel Calibration of Accident/Remote Shutdown Monitoring Auxiliary Feedwater Flow Loop 1-LPF-3-147B, Rev. 5
| |
| : AOI-35, Loss of Offsite Power, Rev. 34
| |
| : 0-SI-82-11-A, Monthly Diesel Generator Start and Load Test 1 A-A, Rev. 24
| |
| : AOI-8, Tornado Watch or Warning, Rev. 31
| |
| : N3-63-4001, Safety Injection System
| |
| : 1-SI-0-902, Testing Setpoint of Safety Relief Valves ASME Section XI Category "C" Valves, Rev. 20
| |
| : CM Chapter 4.08, Non-Oxidizing Biocide Injection into the AFW ERCW B Supply Line for Control of Asiatic Clams, Zebra Mussels, Rev. 3
| |
| : 0-SI-67-902-A, Essential Raw Cooling Water Pump B-A and Pump D-A Performance Test
| |
| (Most Recent Revision, Not Yet Performed), Rev. 22
| |
| : MI-67.001, Maintenance Instruction, Removal, Inspection and Repair of Essential Raw Cooling Water Pumps, Rev. 18Test ProceduresEPT-163, Generic Letter 89-13 Inspections (RAW Water Systems and Local Area Air Handler
| |
| : Inspection and Documentation); B ESW Pump Discharge Strainer, 4/14/2006; A ESW Pump Discharge Strainer, Dated 4/23/2006
| |
| : AB-31D-01, HVAC Air Balance Test for Shutdown Board Room Ventilation System, Dated
| |
| : 9/14/95
| |
| : 03-013478, Adjust Impeller Clearance of ERCW Pump G-B in support of Procedure 0-SI-67-
| |
| : 901-B, Dated 12/18/03
| |
| : 05-815841, Chiller A-A, Annual Inspection of SD BD RM CHLR PKGS, Dated 05/17/06
| |
| : 4Attachment05-823341, Inspect SHTDN BD RM CHILL A-A, Dated 10/31/0606-810943, Chiller A-A, SDBD Room Chiller Refrigerant Sampling and Clean Up Kit Inspection/Installation of Filter, Dated 01/24/07
| |
| : 06-810942, Chiller A-A, SDBR Chiller Oil Filter Replacement, Dated 01/25/07
| |
| : 05-882998, Chiller B-B, Annual Inspection of SD BD RM CHLR PKGS, Dated 10/16/06
| |
| : 05-822500, Chiller B-B, Room Chiller Refrigerant Sampling and Clean Up Kit Inspection/Installation of Filter, Dated 09/28/06
| |
| : 06-818617, 0-SI-31-904-B, Shutdown Board Room Chiller Water Circulation Pump B-B
| |
| : Quarterly Performance Test, Dated 01/04/07
| |
| : 06-810156, Chiller B-B, Room Chiller Refrigerant Sampling and Clean Up Kit Inspection/Installation of Filter, Dated 02/01/07
| |
| : 1-SI-63-905, Boron Injection Check Valve Flow Test During Refueling Outages, Dated 3/23/99,
| |
| : 9/28/99, 9/28/00, 3/12/02, 3/12/02,
| |
| : 3/21/05, and 11/05/06
| |
| : 1-SI-62-901-B, Centrifugal Charging Pump 1B-B Quarterly Performance Test, Dated 9/16/05
| |
| and11/13/07
| |
| : 1-SI-62-901-A, Centrifugal Charging Pump 1A-A Quarterly Performance Test, Dated 11/17/06
| |
| and 1/24/07
| |
| : 0-SI-67-901-A, Essential Raw Cooling Water Pump A-A and Pump C-A Performance Test, Dated 1/24/05, 4/20/05, and 10/20/06
| |
| : 0-SI-67-901-B, Essential Raw Cooling Water Pump E-B and Pump G-B Performance Test, Dated 1/19/00,12/06/03, 9/28/06, and 12/11/06
| |
| : 0-SI-67-902-A, Essential Raw Cooling Water Pump B-A and Pump D-A Performance Test, Dated 10/29/06 and 12/31/06
| |
| : 0-SI-67-902-B, Essential Raw Cooling Water Pump F-B and Pump H-B Performance Test, Dated 9/28/06 and 12/18/06
| |
| : 0-PMP-040-0065MH1, 1E Manhole and Sump Inspection, Rev. 9
| |
| : 0-SI-82-3, 18 Month Loss of Offsite Power with Safety Injection Test, Appendices I, L, & P, performed 11/12/06, Rev. 33
| |
| : 0-SI-215-21-A, Diesel Generator 1A-A Battery Quarterly Inspection, Rev. 4
| |
| : 0-SI-215-1, Diesel Generator Battery Weekly Inspection, Rev. 5
| |
| : 0-SI-215-31-A, Diesel Generator 1A-A Battery Annual Inspection, Rev. 4
| |
| : 0-SI-215-41-A, Diesel Generator 1A-A 18 Month Service Test and Battery Charger Test, Rev. 7
| |
| : 0-SI-215-51-A, Diesel Generator 1A-A 60 Month Performance Test and Battery Charger Test, Rev. 4
| |
| : MI-57.001, 6900 Volt Circuit Breaker Inspection, Rev. 29
| |
| : MI-57.002, Westinghouse DS Circuit Breaker Routine Maintenance, Inspection, and Testing, Rev. 36
| |
| : MI-57.006, 6900 Volt Circuit Breaker Overhaul Inspection, Rev. 4
| |
| : TI-109, Breaker Testing and Maintenance Program, Rev. 4
| |
| : TI-67.001, Component Flow Blockage Testing - Essential Raw Cooling Water, (Train A), Rev. 14
| |
| : TI-100.001, Inservice Testing of Pumps, Rev. 6Design Changes/ModificationsDCN-52179, Shutdown Board Room A and B Air Handling Unit A-A Motor Replacement and
| |
| : Circuit Protective Devices Settings Changes, Rev. A
| |
| : 5AttachmentDCN S-31717-A, Revised SDBD Room Chiller Start Time on Loss of Power, Dated 6/29/94DCN 51370A, Vital Inverter Replacement [scope, design criteria, technical evaluation, interfaces, 10
| |
| : CFR 50.59 screening], Dated 11/4/04Design Basis DocumentsN3-3B-4002, Auxiliary Feedwater System Description, Rev. 12N3-32-4002, Compressed Air System Description, Rev. 7
| |
| : N3-30AB-4001, Auxiliary Building Heating, Ventilation, and Air Conditioning System Description, Rev. 24
| |
| : AOP-020-BD, Loss of RCS Inventory or RHR While Shutdown - Basis Document, Rev. 1Problem Event Reports (PERs)22091, ERCW Pump G-B Did Not Meet Acceptance Criteria of Performance Test 0-SI-67-902-B88809, ERCW Pump A-A in Alert Range
| |
| : 80873, ERCW Pump C-A Failed SI Test, 0-SI-67-901-A
| |
| : 74180, SDBD Room Chiller A Tripped on High Oil/Disch Temp
| |
| : 106173, SDBD Rm Chiller B Tripped on 6/28/06
| |
| : 106175, Shutdown Board Room Chiller A Trip Possible Air in Sensing Line
| |
| : 2750, INPO Identified Two Issues with IST for ERCW Pumps, (Instrument Uncertainty and River Water Temperatures Impacts)
| |
| : 110908, SDBR Chiller "A" Tripped During Performance of SOI-67.01
| |
| : 209, Agastat Relay Replacement
| |
| : 99316, Load Shed Logic Panel Relay SD BD 1-A
| |
| : 103208, 10 CFR Part 21 - Ametek SCI Inverter Capacitors
| |
| : 107266, GE Relay Paddle
| |
| : 111392, Incomplete Breaker Test Data
| |
| : 115951, Diesel Generator Frequency Safety Analysis Interaction
| |
| : 2195, K1 Contactor Failure in EDG at Palo Verde
| |
| : 141831, Degraded Tube Steel Brace in ESW Strainer BasketWork Orders01-003615-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (EDG output breaker), Dated 7/29/02
| |
| : 01-014550-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (ERCW
| |
| pump motor feeder breaker), Dated 5/10/02
| |
| : 01-016659-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (ERCW
| |
| pump motor feeder breaker), Dated 8/2/02
| |
| : 2-001714-000, 6900 V Circuit Breaker Inspection - Perform
| |
| : MI-57.001 (ERCW pump motor feeder breaker), Dated 8/14/02
| |
| : 2-012941-000, Periodic Calibration, 1-LPF-70-81A-A, Thermal Barrier Return Header Flow , Dated 9/25/0302-012942-000, Periodic Calibration, 1-LPF-070-0081E/A-A Thermal Barrier Supply Header Flow Differential , Dated 9/28/0302-012961-000, Periodic Calibration, 1-LPF-070-0081B/D-B Thermal Barrier Supply Header
| |
| : 6Attachment Flow Differential , Dated 9/26/0302-016617-000, Megger 1E 6900 V Cables in Manholes (ERCW pump A-A), Dated 2/19/03
| |
| : 03-006029-000, Clean, Inspect, and Restore 1E Manhole / Sumps , Dated 1/20/04
| |
| : 03-010639-000, 6900 V Circuit Breaker Inspection - Perform
| |
| : MI-57.001 (ERCW pump motor feeder breaker), Dated 8/30/04
| |
| : 03-011258-000, Clean, Inspect, and Restore 1E Manhole / Sumps , Dated 6/24/04
| |
| : 03-012186-000, Megger 1E 6900 V Cables in Manholes (EDG 1A-A), Dated 3/15/04
| |
| : 03-012196-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (ERCW
| |
| pump motor feeder breaker), Dated 11/12/04
| |
| : 03-014339-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (EDG
| |
| output breaker), Dated 10/20/04
| |
| : 03-014640-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (CCP
| |
| motor feeder breaker), Dated 5/27/05
| |
| : 03-018612-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (EDG
| |
| output breaker), Dated 5/21/05
| |
| : 03-018699-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (CCP
| |
| motor feeder breaker), Dated 9/27/04
| |
| : 03-018703-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 8/1/05
| |
| : 03-021591-000, Periodic Calibration, 1-LPF-70-81A-A, Thermal Barrier Return Header Flow, Dated 3/8/0503-021592-000, Periodic Calibration, 1-LPF-070-0081E/A-A Thermal Barrier Supply Header Flow Differential, Dated 3/8/0503-021611-000, Periodic Calibration, 1-LPF-070-0081B/D-B Thermal Barrier Supply Header Flow Differential, Dated 3/21/0504-810576-000, Megger 1E 6900 V Cables in Manholes (ERCW pump A-A), Dated 11/8/04
| |
| : 04-810634-000, Clean, Inspect, and Restore 1E Manhole / Sumps, Dated 4/22/05
| |
| : 04-811232-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 8/8/05
| |
| : 04-811382-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 8/28/05
| |
| : 04-811453-000, Megger 1E 6900 V Cables in Manholes (ERCW pump A-A), Dated 2/14/05
| |
| : 04-811860-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 10/7/05
| |
| : 04-812769-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 10/13/05
| |
| : 04-813557-000, Megger 1E 6900 V Cables in Manholes (EDG 2A-A), Dated 5/27/05
| |
| : 04-813566-000, Megger 1E 6900 V Cables in Manholes (ERCW pump C-A), Dated 1/23/05
| |
| : 04-813732-000, Megger 1E 6900 V Cables in Manholes (EDG 1B-B), Dated 3/16/05
| |
| : 04-813739-000, Diesel Generator 1B-B Calibration, Solid State Speed Sensing, Dated 3/14/05
| |
| : 04-813956-000, Megger 1E 6900 V Cables in Manholes (EDG 2B-B), Dated 3/16/05
| |
| : 04-813963-000, Diesel Generator 2B-B Calibration, Solid State Speed Sensing , Dated 3/15/05
| |
| : 04-814704-000, Diesel Generator 1A-A Calibration, Solid State Speed Sensing , Dated 5/10/05
| |
| : 04-815021-000, Megger 1E 6900 V Cables in Manholes (EDG 1A-A), Dated 5/10/05
| |
| : 04-816376-000, 6900 V Circuit Breaker Inspection - Perform
| |
| : MI-57.001 (EDG output breaker), Dated 4/17/05
| |
| : 04-816593-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass
| |
| : 7Attachment breaker), Dated 9/25/0504-816628-000, Clean, Inspect, and Restore 1E Manhole / Sumps , Dated 10/21/05
| |
| : 04-816933-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 7/10/05
| |
| : 04-819751-000, 480 V Breaker Inspection - Perform
| |
| : MI-57.002 (current limiting reactor bypass breaker), Dated 7/7/05
| |
| : 04-824791-000, Clean, Inspect, and Restore 1E Manhole / Sumps , Dated 4/18/06
| |
| : 05-811033-000, Megger 1E 6900 V Cables in Manholes (EDG 1A-A), Dated 5/8/06
| |
| : 05-812455-000, 6900 V Circuit Breaker Inspection and Overhaul - Perform
| |
| : MI-57.006 (ERCW
| |
| pump motor feeder breaker), Dated 6/16/06
| |
| : 05-815120-000, Periodic Calibration, 1-LPF-70-81A-A, Thermal Barrier Return Header Flow, Dated 9/16/0605-815121-000, Periodic Calibration, 1-LPF-070-0081E/A-A Thermal Barrier Supply Header Flow Differential, Dated 9/22/0605-815139-000, Periodic Calibration, 1-LPF-070-0081B/D-B Thermal Barrier Supply Header Flow Differential, Dated 9/25/0605-820019-000, Diesel Generator 1A-A 18 Month Service Test and Battery Charger Test,
| |
| : Dated 5/9/06
| |
| : 05-821048-000, Clean, Inspect, and Restore 1E Manhole / Sumps, Dated 1/8/07
| |
| : 05-821053-000, Megger 1E 6900 V Cables in Manholes (ERCW pump D-A), Dated 1/26/07
| |
| : 05-821295-000, Megger 1E 6900 V Cables in Manholes (ERCW pump E-B), Dated 9/19/06
| |
| : 05-821964-000, Megger 1E 6900 V Cables in Manholes (ERCW pump B-A), Dated 10/20/06
| |
| : 05-823-669-000, Calibration / test of EDG 1B-B breaker voltage permissive, Dated 9/21/06
| |
| : 05-823670-000, Calibration / test of EDG 1A-A breaker voltage permissive, Dated 12/18/06
| |
| : 05-823786-000, Calibration / test of EDG 2A-A breaker voltage permissive, Dated 12/28/06
| |
| : 05-824272-000, 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test, Dated 5/19/06
| |
| : 05-825388-000, 18 Month B Train Motor Operated Valve Thermal Overload Relay Bypass Circuit Functional Test, Attachment C, 1-FCV-70-87-B, Dated 8/25/06
| |
| : 06-803731-000, Diesel Generator 1A-A 60 Month Performance Test and Battery Charger Test (vendor test), Dated 4/4/06
| |
| : 06-810312-000, 18 Month A Train Motor Operated Valve Thermal Overload Relay Bypass Circuit Functional Test, Attachment C, 1-FCV-70-133-A, Dated 9/6/06
| |
| : 06-811325-000, 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test, Dated 7/11/06
| |
| : 06-810637-000, Diesel Generator 1A-A Battery Annual Inspection, Dated 5/8/0606-814573-000, 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test, Dated 11/22/06
| |
| : 06-817789-000, Diesel Generator 1A-A Battery Quarterly Inspection, Dated 12/18/06
| |
| : 06-820061-000, 1A-A Battery Has an Intermittent -110 Volt Ground, Dated 10/4/06
| |
| : 06-820876-000, Diesel Generator Battery Weekly Inspection, Dated 2/26/0705-820645-000 Train B Igniter Test Data, Dated 11/19/2006
| |
| : 05-820744-000 Train B Igniter Test Data, Dated 11/19/2006
| |
| : 04-815848-000,Train A Igniter Test Data, Dated 3/23/2005
| |
| : 04-815938-000,Train A Igniter Test Data, Dated 3/23/2005
| |
| : 2-013209-000, CCX HX C Eddy Current Testing and Inspection, Dated 11/13/2003
| |
| : 03-021246-000, CCS HX A Eddy Current Testing and Inspection, Dated 3/7/2005
| |
| : 8Attachment05-815458-000, CCX HX C Eddy Current Testing and Inspection, Dated 9/22/200604-816077-000, CCS HX A ERCW Side Flow and D/P Monitoring, Dated 3/21/2005
| |
| : 04-823716-000, CCS HX A ERCW Side Flow and D/P Monitoring, Dated 1/27/2005
| |
| : 05-820844-000, CCS HX C ERCW Side Flow and D/P Monitoring, Dated 8/14/2006
| |
| : 04-816025-000, CCS HX C ERCW Side Flow and D/P Monitoring, Dated 2/9/2005
| |
| : 04-811997-000, DG 2A1 Inspection of Jacket Water Cooler, Dated 5/19/2005
| |
| : 05-815800-000, DG 2A1 Inspection of Jacket Water Cooler, Dated 4/17/2006
| |
| : 04-816985-000, DG 1A1 Inspection of Jacket Water Cooler, Dated 5/11/2005
| |
| : 04-816986-000, DG 1A2 Inspection of Jacket Water Cooler, Dated 5/11/2005
| |
| : 04-812636-000, DG 1B1 Inspection of Jacket Water Cooler, Dated 3/13/2005
| |
| : 04-812637-000, DG 1B2 Inspection of Jacket Water Cooler, Dated 3/13/2005
| |
| : 04-811998-000, DG 2A2 Inspection of Jacket Water Cooler, Dated 5/18/2005
| |
| : 05-815816-000, DG 2A2 Inspection of Jacket Water Cooler, Dated 4/17/2006
| |
| : 04-812806-000, DG 2B1 Inspection of Jacket Water Cooler, Dated 3/14/2005
| |
| : 05-812438-000, DG 2B1 Inspection of Jacket Water Cooler, Dated 5/11/2006
| |
| : 04-812807-000, DG 2B2 Inspection of Jacket Water Cooler, Dated 3/15/2005
| |
| : 05-812439-000, DG 2B2 Inspection of Jacket Water Cooler, Dated 5/1/2006
| |
| : 03-000492-000, Main Control Room Chiller A-A HX Inspection, Dated 4/30/2003
| |
| : 03-015726-000, Main Control Room Chiller A-A HX Inspection, Dated 6/6/2005
| |
| : 04-820460-000, Main Control Room Chiller A-A HX Inspection, Dated 6/15/2005
| |
| : 04-812811-000, Main Control Room Chiller B-B HX Inspection, Dated 12/15/2004
| |
| : 05-821313-000, Main Control Room Chiller B-B HX Inspection, Dated 10/2/2006
| |
| : 03-018581-000, Shutdown Board Room Chiller A-A Inspection, Dated 5/19/2004
| |
| : 05-817681-000, Shutdown Board Room Chiller A-A Inspection, Dated 5/16/2006
| |
| : 03-012647-000, Shutdown Board Room Chiller B-B Inspection, Dated 9/15/2004Drawings1-47W845-1, Mechanical Flow Diagram - Essential Raw Cooling Water System, Rev. 501-47W866-8, Flow Diagram Heating Cooling & Ventilation Air Flow, Rev. 20
| |
| : 1-47W611-65-1, Electrical Logic Diagram Emergency Gas Treatment System, Rev. 8
| |
| : 1-45W760-30-20, Wiring Diagrams Ventilating System Schematic Diagrams, Rev. 9
| |
| : 1-47W610-30-5, Electrical Control Diagram Containment Ventilating System, Rev. 2
| |
| : 47W450-3D, Mechanical Essential Raw Cooling Water, Rev. H
| |
| : 17W302-1, Mechanical Essential Raw Cooling Water, Control Air & HPFP Piping, Rev. 17
| |
| : 1-47W866-9, Flow and Control Diagrams Heating Ventilating Air Flow, Rev. 16
| |
| : 1-47W611-30-7, Electrical Logic Diagram Ventilation System, Rev. 9
| |
| : 17W910-1 RL, Diesel Generator Building Mechanical Heating & Ventilation
| |
| : 17W910-2 RJ, Diesel Generator Building Mechanical Heating & Ventilation
| |
| : 17W910-4 RE, Diesel Generator Building Mechanical Heating & Ventilation
| |
| : 1-47W811-1, Flow Diagram Safety Injection System, Rev. 44
| |
| : 1-47W813-1, Flow Diagram Reactor Coolant System, Rev. 41
| |
| : 1-47W845-2, Mechanical Flow Diagram Essential Raw Cooling Water System, Rev. 63
| |
| : 1-47W610-67-5, Electrical Control Diagram ERCW System, Rev. 14
| |
| : 1-47W611-67-5, Electrical Logic Diagram Essential Raw Cooling Water, Rev. 10
| |
| : 54A0241, Air Operated Valve (FCV-62-93), Rev. B
| |
| : 1-45W760-30-18, Wiring Diagrams Ventilating System Schematic Diagram, Rev. 10
| |
| : 9Attachment47E235-07, Environmental Data Environment - Mild
| |
| : EL 757.0, Rev. 51-47W865-8, Flow Diagram Air Conditioning Chilled Water (Aux Bldg), Rev.18
| |
| : 1-47W866-3, Flow Diagram Heating Ventilation & Air Cond Air Flow (Aux Bldg), Rev. 37
| |
| : 47W920-35, Mechanical Heating, Ventilating and Air Conditioning (Aux Bldg), Rev. 30
| |
| : 47W920-8, Mechanical Heating, Ventilation and Air Conditioning, Rev. 41
| |
| : 1-47W811-1, Flow Diagram Safety Injection System, Rev. 44
| |
| : 1-47W610-62-2, Electrical Control Diagram Chemical & Volume Control System, Rev. 23
| |
| : E2-56039, Refueling Water Storage Tank General Arrangement, Dated 04/22/77
| |
| : E10-56039, Refueling Water Storage Tank Vortex Suppressor, Dated 04/22/77
| |
| : E11-56039, Refueling Water Storage Tank Roof Manhole & Roof Vent, Dated 04/22/77
| |
| : 15W810-1, Conduit & Grounding General Plan, Rev. 27
| |
| : 1-15E500-2, Key Diagram, Station Auxiliary Power System, Rev. 33
| |
| : 1-45W600-57-19, Separation & Miscellaneous Auxiliary Relays Schematic Diagrams [EDG
| |
| breaker], Rev. 16
| |
| : 1-45W600-57-23, Separation & Miscellaneous Auxiliary Relays Schematic Diagram [EDG
| |
| breaker], Rev. 6
| |
| : 1-45W700-1, Key Diagram, 120 Vac & 125 Vdc Vital Plant Control Power System, Rev. 24
| |
| : 1-45W760-82-2, Standby Diesel Generator System Schematic Diagrams DG 1A-A [EDG
| |
| breaker], Rev. 17
| |
| : 1-45W760-82-3, Standby Diesel Generator System Schematic Diagrams DG 1A-A [EDG
| |
| breaker], Rev. 10
| |
| : 1-45W760-82-4, Standby Diesel Generator System Schematic Diagrams DG 1A-A [EDG
| |
| breaker], Rev. 19
| |
| : 1-45W760-82-5,Standby Diesel Generator System Schematic Diagrams DG 1A-A [EDG
| |
| breaker], Rev. 19
| |
| : 1-45W760-82-6, Standby Diesel Generator System Schematic Diagrams DG 1A-A [EDG
| |
| breaker], Rev. 22
| |
| : 1-45W760-211-4, 6900 V Shutdown Power Schematic Diagram [EDG breaker], Rev. 15
| |
| : 1-45W760-212-2, 480V Shutdown Power Schematic Diagrams [reactor bypass breaker], Rev. 12
| |
| : 1-45W760-212-4, 480V Shutdown Power Schematic Diagrams [reactor bypass breaker], Rev. 14
| |
| : 1-45W760-268-1, Permanent Hydrogen Mitigation System Schematic Diagrams, Rev. 6
| |
| : 1-47W600-825, Electrical Instrument Sensing Line Slope Configuration - Interface, Rev. 0
| |
| : 1-47W600-829, Electrical Instrument Sensing Line Slope Configuration - Interface, Rev. 0
| |
| : 1-47W610-70-3, Electrical Control Diagram, Component Cooling Water System , Rev. 16
| |
| : 1-47W760-70-4, Component Cooling Water System Schematic Diagrams [1-FCV-70-87-B, 1-
| |
| : FCV-70-90-A], Rev. 10
| |
| : 1-47W760-70-5, Component Cooling Water System Schematic Diagrams [1-FCV-70-133, 1-
| |
| : FCV-70-134], Rev. 11
| |
| : 1-47W760-211-4, 6900 Vac Shutdown Power Schematic Diagram [EDG feeder breakers], Rev. 15
| |
| : 1-47W760-270-2, Miscellaneous System Schematic Diagram [thermal overload bypass control], Rev. 25
| |
| : 1-75W550, Miscellaneous Control Circuits, 250 Vdc & 120 Vac Schematics [EDG breaker], Rev. 10
| |
| : 10Attachment1-75W551, Miscellaneous Control Circuits, 250 Vdc & 120 Vac Schematics [EDG breaker],
| |
| : Rev. 3
| |
| : 45N700-3, Key Diagram, Permanent Hydrogen Mitigation Power System, Rev. 0
| |
| : 47E235-41, Environmental Data, Upper Compartment, Rev. 11
| |
| : 47E235-42, Environmental Data, Lower Compartment, Rev. 11
| |
| : 47W600-132, Electrical Instruments and Controls [thermal barrier flow instrument installation details], Rev. 12
| |
| : C379C11501,Schematic Diagram, DC Distribution Panel, Diesel Generator System, Rev. 905
| |
| : D8061727C, Bailey Elementary Diagram [component cooling water thermal barrier flow]
| |
| : 1-45W706-4, Wiring Diagram 120 VAC Vital Instrument Power Boards I-IV and
| |
| : II-IV Connection Diagrams, Rev. 46
| |
| : 1-45W706-10, Wiring Diagram 120 VAC Vital Instrument Power Boards I-IV Instrument Loading Sheet 6, Rev. 3Miscellaneous DocumentsN3-30DB-4002, System Description for Diesel Generator Building Ventilation System, Rev. 13N3-67-4002, System Description for Essential Raw Cooling Water System, Rev. 19
| |
| : Crispin Valves "P Series" Pressure Air Release Valve for Normal Service, Rev. 0DCN 52011-A, Install Pressure Air Release Valves on ERCW Discharge Headers "A" & "B"
| |
| : ASME Operations and Maintenance Standard
| |
| : OM-1, 1987 Edition
| |
| : T69
| |
| : 070108 131, ERCW Cement-Mortar Lined Piping, Dated 1/8/2007
| |
| : WBPLMN-97-102-0,
| |
| : STI-97-01 Safety Assessment/ Evaluation Functional Evaluation for PERs 95337 &
| |
| : 102939, Rev. 4
| |
| : G-94, Piping Installation, Modification and Maintenance, Rev. 2
| |
| : T04
| |
| : 940304 853, TVA Response to Generic Letter 89-13, Dated 3/4/1994
| |
| : T04
| |
| : 940523 909, TVA Response to Generic Letter 89-13, Dated 5/23/1994
| |
| : T04
| |
| : 979730 464, TVA Response to Generic Letter 89-13, Dated 7/30/1997
| |
| : T04
| |
| : 971105 495, TVA Response to Generic Letter 89-13, Dated 11/5/1997
| |
| : WAT-D-8902, Watts Bar Re-Verification Program Safety Injection System Reissuance, Rev. 1
| |
| : WBN Precautions Limitation and Setpoints, Rev. 9
| |
| : N3-67-4002, System Description for Essential Raw Cooling Water System 67, Rev. 19
| |
| : 73C34-83577, Control Valve Specification (for
| |
| : FCV-67-188), Dated 02/02/77
| |
| : WAT-D-3702, Westinghouse Letter to TVA, ECCS Injection to Recirculation Switchover, Dated
| |
| : 08/15/79
| |
| : W10A52 C-K, Sequoyah and Watts Bar Nuclear Plants Unit 1 and 2 - Vortex Test of the Refueling Water Storage Tank, Dated 11/19/76
| |
| : No.
| |
| : 1000067158, TVA Central Laboratories Services, Test Report for Analog Pressure Gauge, Heise, Dated 2/07/02
| |
| : WAT-D-9662, Westinghouse Letter to
| |
| : TVA-WBN, CCP, RHR and SI Pumps Vibration Criteria, Dated 3/11/94
| |
| : WAT-D-11294, Westinghouse Letter to WBN Regarding CCP Runout Evaluation, Dated 1/13/06
| |
| : WAT-D-10872, Westinghouse Letter to WBN Regarding CCP Performance, Dated 10/25/00
| |
| : 74C63-83090, Diesel Engine-Driven Emergency Generator Power Packages, Guaranteed Contract Data, Equipment Data, Dated 10/19/73
| |
| : NER-020446, Disposition of NRC
| |
| : IN 2002-12: Submerged Safety-Related Cables, Dated 6/18/02
| |
| : 11AttachmentNER-970399, Disposition of NRC
| |
| : IN 97-08: Potential Failures of General Electric Magne-BlastŽ
| |
| : Circuit Breaker Subcomponents, Dated 7/12/99
| |
| : NER-980795, Disposition of IE Notice 98-38: Metal-Clad Circuit Breaker Maintenance Issues Identified by NRC Inspections, Dated 11/19/98
| |
| : PTI-262-01, Integrated Safeguards Test, Section 6.7, Data Sheet 8.31, SER Timing & Data Sheet 8.39, Transient Load Response, Dated 11/25/94
| |
| : Westinghouse Nuclear Safety Analysis Letter
| |
| : NSAL-05-03, CCP Runout During SI, Dated 5/4/2005Corrective Action documents initiated due to CDBI activity
| |
| :PER
| |
| : 119852, Drawing 1-45W706-8 is incorrect.
| |
| : 1-LI-3-43B and 1-LI-3-111B should be listed as loads for breaker 37 on board 1-III.
| |
| : PER 119716, Document that an LER was prepared for 95337
| |
| : PER 118771, Non-ASME Code Air Release Valves
| |
| : PER 120005, DGB Ventilation During Tornado Warning
| |
| : PER 120741, Documentation error in calculation WBNOSG4-136
| |
| : PER 120010, Error in Calculation
| |
| : EPM-JFL-060395 ERCW Pump Lift Clearances
| |
| : PER 120696, Flow/dp Error in Table A of
| |
| : TI-67.001, Component Flow Blockage Testing -
| |
| : ERCW
| |
| : PER 120746, ERCW Pump Calculation Error in procedure 0-SI-67-901-B
| |
| : PER 120979, ECCS Pump NPSH Calc Affected by dP across RWST Vent
| |
| : PER 120933, UFSAR Error NPSH Calculation Takes Credit for RWST Water
| |
| : PER 121023, Calculation Errors in EPM-CES-092689
| |
| : PER 119810, Two large wrenches were lying unrestrained on the diesel fuel oil transfer pump skid
| |
| : PER 121032, Incorrect documentation of M&TE usage range in speed switch calibration
| |
| : PER 121034, Inspection record for 6900 Vac circuit breaker Megger test showed 1400
| |
| megohms C to A phase where data was initially recorded but 14000 megohms in trending log;
| |
| acceptance criterion is 2000 meghoms; apparent typographical error.
| |
| : WO 07-812580-000, Small puddle of fuel oil on the floor of the fuel oil transfer pump room -
| |
| clean up, investigate and repair
| |
| : WO 07-812579-000, Indicating light lens for control switch located on diesel generator air dryer station found broken - repair
| |
| }} | | }} |
|
---|
Category:Inspection Report
MONTHYEARIR 05000390/20240022024-08-0707 August 2024 Integrated Inspection Report 05000390/2024002 and 05000391/2024002 Rev IR 05000390/20244402024-07-12012 July 2024 95001 Supplemental Inspection Supplemental Report 05000390-2024440 and 05000391-2024440 and Follow-Up Assessment Letter IR 05000390/20240012024-05-14014 May 2024 Integrated Inspection Report 05000390/2024001 and 05000391/2024001 IR 05000391/20240072024-04-30030 April 2024 Assessment Follow-up Letter for Watts Bar Nuclear Plant, Unit 2 – Report 05000391/2024007 IR 05000390/20244012024-04-0202 April 2024 – Security Baseline Inspection Report 05000390/2024401 and 05000391/2024401 - (Public) IR 05000390/20240102024-03-0808 March 2024 Age-Related Degradation Inspection Report 05000390/2024010 and 05000391/2024010 IR 05000390/20230062024-02-28028 February 2024 Annual Assessment Letter for Watts Bar Nuclear Plant Units 1 and 2 - Report 05000390/2023006 and 05000391/2023006 IR 05000390/20230042024-02-13013 February 2024 Integrated Inspection Report 05000390/2023004 and 05000391/2023004 and Apparent Violation IR 05000390/20234412023-12-21021 December 2023 Plantfinal Significance Determination for a Security-Related Greater than Green Finding, Nov, and Assessment Follow-up, 05000390-2023441 and 05000391-2023441-Public IR 05000390/20234042023-12-14014 December 2023 Security Baseline Inspection Report 05000390/2023404 and 05000391/2023404 IR 05000390/20230102023-11-30030 November 2023 RE-Issue Watts Bar Nuclear Plant - Biennial Problem Identification and Resolution Inspection Report 050000390/2023010 and 05000391/2023010 and Apparent Violation IR 05000390/20230032023-11-13013 November 2023 Integrated Inspection Report 05000390/2023003 and 05000391/2023003 and Apparent Violation ML23296A0242023-10-24024 October 2023 Biennial Problem Identification and Resolution Inspection Report and Preliminary Greater than Green Finding and Apparent Violation IR 05000390/20230052023-08-30030 August 2023 Updated Inspection Plan for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2023005 and 05000391/2023005 IR 05000390/20230022023-08-16016 August 2023 Reissue - Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2023002 and 05000391/2023002 ML23220A1582023-08-0909 August 2023 Integrated Inspection Report 05000390/2023002 and 05000391/2023002 IR 05000390/20230112023-07-24024 July 2023 Quadrennial Focused Engineering Inspection (FEI) Commercial Grade Dedication Report 05000390 2023011 and 05000391 2023011 IR 05000390/20234032023-05-30030 May 2023 Cyber Security Inspection Report 05000390/2023403 and 05000391/2023403 IR 05000390/20220032023-05-0909 May 2023 Reissue Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2022003 and 05000391/2022003 IR 05000390/20230012023-05-0404 May 2023 Integrated Inspection Report 05000390/2023001 and 05000391/2023001 IR 05000390/20234012023-03-13013 March 2023 Security Baseline Inspection Report 05000390 2023401 and 05000391 2023401 IR 05000390/20220062023-03-0101 March 2023 Annual Assessment Letter for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2022006 and 05000391/2022006 IR 05000390/20220042023-02-10010 February 2023 Integrated Inspection Report 05000390/2022004 and 05000391/2022004 IR 05000390/20224202022-12-0101 December 2022 Security Baseline Inspection 05000390/2022420 and 05000391/2022420 Cover Letter ML22318A0072022-11-14014 November 2022 Integrated Inspection Report 05000390/2022003 and 05000391/2022003 IR 05000390/20220102022-10-28028 October 2022 Design Basis Assurance Inspection (Teams) Inspection Report 05000390/2022010 and 05000391/2022010 IR 05000390/20223012022-09-30030 September 2022 NRC Operator License Examination Report 05000390/2022301 and 050000391/2022301 ML22256A2952022-09-14014 September 2022 Operation of an Independent Spent Fuel Storage Installation Report 07201048/2022001 IR 05000390/20220052022-08-31031 August 2022 Updated Inspection Plan for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2022005 and 05000391/2022005 - Final IR 05000390/20220022022-08-12012 August 2022 Integrated Inspection Report 05000390/2022002 and 05000391/2022002 and Exercise of Enforcement Discretion IR 05000390/20224012022-08-0303 August 2022 Security Baseline Inspection Report 05000390/2022401 and 05000391/2022401 IR 05000390/20220012022-05-11011 May 2022 Integrated Inspection Report 05000390/2022001 and 05000391/2022001 ML22123A2412022-05-11011 May 2022 Review of the Fall 2021 Mid-Cycle Steam Generator Tube Inspection Report IR 05000390/20210062022-03-0202 March 2022 Annual Assessment Letter for Watts Bar Nuclear Plant, Units 1 and 2 (Report No. 05000390/2021006 and 05000391/2021006) IR 05000390/20210042022-02-10010 February 2022 Integrated Inspection Report 05000390/2021004 and 05000391/2021004 IR 05000390/20210032021-11-10010 November 2021 Integrated Inspection Report 05000390/2021003 and 05000391/2021003 IR 05000390/20214022021-11-0202 November 2021 Security Baseline Inspection Report 05000390/2021402 and 05000391/2021402 IR 05000390/20214032021-10-26026 October 2021 Material Control and Accounting Program Inspection Report 05000390/2021403 and 05000391/2021403 (OUO Removed) ML21263A0042021-09-24024 September 2021 Review of the Fall 2020 Steam Generator Tube Inspection Report IR 05000390/20214012021-08-31031 August 2021 Security Baseline Inspection Report 05000390/2021401 and 05000391/2021401 IR 05000390/20210052021-08-24024 August 2021 Updated Inspection Plan for Watts Bar Nuclear Plant, Units 1 & 2 Report 05000390/2021005 and 05000391/2021005 IR 05000390/20210022021-08-0404 August 2021 Integrated Inspection Report 05000390/2021002 and 05000391/2021002 IR 05000390/20210122021-07-0101 July 2021 Biennial Problem Identification and Resolution Inspection Report 05000390/2021012 and 05000391/2021012 IR 05000390/20210102021-05-0606 May 2021 NRC Inspection Report 05000390/2021010 and 05000391/2021010 IR 05000390/20210012021-05-0505 May 2021 Integrated Inspection Report 05000390/2021001 and 05000391/2021001 IR 05000390/20210112021-04-13013 April 2021 Triennial Fire Protection Inspection Report 05000390/2021011 and 05000391/2021011 IR 05000390/20200062021-03-0303 March 2021 Annual Assessment Letter for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2020006 and 05000391/2020006 IR 05000390/20200042021-02-11011 February 2021 Integrated Inspection Report 05000390/2020004, 05000391/2020004, 07201048/2020002, and Exercise of Enforcement Discretion IR 05000390/20203012021-01-19019 January 2021 Reissue - Watts Bar Nuclear Plant - NRC Operator License Examination Report 05000390/2020301 and 05000391/2020301 IR 05000390/20204032020-12-15015 December 2020 Security Baseline Inspection Report 05000390/2020403 and 05000391/2020403 2024-08-07
[Table view] Category:Letter
MONTHYEARML24219A0262024-08-12012 August 2024 Request for Withholding Information from Public Disclosure IR 05000390/20240022024-08-0707 August 2024 Integrated Inspection Report 05000390/2024002 and 05000391/2024002 Rev ML24204A2652024-07-25025 July 2024 Regulatory Audit Summary Related to Request to Revise Technical Specification Surveillance Requirement 3.9.5.1 to Reduce the Residual Heat Removal Flow Rate ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24170A8002024-07-15015 July 2024 Issuance of Amendment Nos. 168 and 74 Regarding Revision to Technical Specification Table 1.1-1 for Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts IR 05000390/20244402024-07-12012 July 2024 95001 Supplemental Inspection Supplemental Report 05000390-2024440 and 05000391-2024440 and Follow-Up Assessment Letter ML24131A0012024-07-0202 July 2024 Issuance of Amendment Nos. 167 and 73 Regarding Adoption of Technical Specification Task Force Traveler TSTF-427-A, Revision 2 CNL-24-052, Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-06-27027 June 2024 Response to Request for Additional Information Regarding Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-24-018, License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS2024-06-25025 June 2024 License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-276-A, Revision 2, Regarding TS 3.8.1 AC Sources – Operating to Clarify Requirements for Diesel Generator Testing (WBN-TS ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc ML24100A7642024-05-16016 May 2024 Issuance of Amendment No. 166 Regarding Revision to Technical Specification 3.8.2, AC Sources-Shutdown, to Remove Reference to C-S Diesel Generator (CNL-23-062) IR 05000390/20240012024-05-14014 May 2024 Integrated Inspection Report 05000390/2024001 and 05000391/2024001 CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000391/20240072024-04-30030 April 2024 Assessment Follow-up Letter for Watts Bar Nuclear Plant, Unit 2 – Report 05000391/2024007 ML24120A1182024-04-29029 April 2024 – Notification of NRC Supplemental Inspection (95001) and Request for Information CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 ML24087A1912024-04-18018 April 2024 Exemption from Select Requirements of 10 CFR Part 73, Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions CNL-24-010, License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19)2024-04-17017 April 2024 License Amendment Request to Recapture Low-Power Testing Time (WBN-TS-23-19) CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update ML24072A0052024-04-15015 April 2024 Issuance of Amendment Nos. 165 and 72 Regarding Increase in the Maximum Number of Tritium Producing Burnable Absorber Rods and Supporting Changes, and Revision to the Updated Final Safety Analysis Report CNL-24-004, Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13)2024-04-0404 April 2024 Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-23-13) IR 05000390/20244012024-04-0202 April 2024 – Security Baseline Inspection Report 05000390/2024401 and 05000391/2024401 - (Public) CNL-24-020, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements2024-04-0101 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Units 1 and 2, Request for Approval of Quality Assurance Program Description and Application to Revise the Technical Specifications Associated with QAPD Requirements CNL-24-008, Guarantee of Payment of Deferred Premiums - 2023 Annual Report2024-03-27027 March 2024 Guarantee of Payment of Deferred Premiums - 2023 Annual Report CNL-24-007, Annual Insurance Status Report2024-03-27027 March 2024 Annual Insurance Status Report CNL-24-025, Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule2024-03-25025 March 2024 Notice of Intent to Pursue License Renewal for Watts Bar Nuclear Plant, Unit 1 - Submittal Schedule ML24081A0262024-03-21021 March 2024 Emergency Plan Implementing Procedure Revisions ML24079A0312024-03-19019 March 2024 Wb 2024-301, Corporate Notification Letter (210-day Ltr) CNL-24-031, Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-03-18018 March 2024 Supplement to Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance CNL-24-029, Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 372024-03-14014 March 2024 Response to Request for Additional Information Regarding Tennessee Valley Authority - Request for Exemption from Specific Requirements of 10 CFR Part 37 CNL-24-028, Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2024-03-14014 March 2024 Response to Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000390/20240102024-03-0808 March 2024 Age-Related Degradation Inspection Report 05000390/2024010 and 05000391/2024010 CNL-24-012, Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance2024-02-28028 February 2024 Request for Exemption from Various Part 72 Regulations Resulting from Fuel Basket Design Control Compliance IR 05000390/20230062024-02-28028 February 2024 Annual Assessment Letter for Watts Bar Nuclear Plant Units 1 and 2 - Report 05000390/2023006 and 05000391/2023006 CNL-24-023, Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 252024-02-20020 February 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revision. Includes CECC EPIP-17, Revision 25 IR 05000390/20230042024-02-13013 February 2024 Integrated Inspection Report 05000390/2023004 and 05000391/2023004 and Apparent Violation ML24038A3342024-02-0707 February 2024 Change to South Texas Project Electric Generating Station (Stepgs) Emergency Plan ML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues ML24008A2462024-01-18018 January 2024 Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-23-052, Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability2024-01-0909 January 2024 Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability CNL-23-062, Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018)2024-01-0808 January 2024 Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018) ML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods CNL-23-069, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000390/20234412023-12-21021 December 2023 Plantfinal Significance Determination for a Security-Related Greater than Green Finding, Nov, and Assessment Follow-up, 05000390-2023441 and 05000391-2023441-Public CNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) 2024-08-07
[Table view] |
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April 20, 2007Tennessee Valley AuthorityATTN:Mr. Preston D. Swafford, ActingChief Nuclear Officer and Executive Vice President6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801SUBJECT:WATTS BAR NUCLEAR PLANT- NRC COMPONENT DESIGN BASISINSPECTION REPORT 05000390/2007006
Dear Mr. Swafford:
On March 9, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection atyour Watts Bar Nuclear Power Plant. The enclosed inspection report documents the inspection findings which were discussed on March 9 and April 18, 2007, with Mr. M. Lorek and other members of your staff. The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.Based on the results of this inspection, the inspectors identified one finding of very low safetysignificance (Green). This finding was determined to involve a violation of NRC requirements.
However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating the finding as a non-cited violation consistent with Section VI.A.1 of the NRC's Enforcement Policy. If you deny this non-cited violation you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Watts Bar Nuclear Power Plant.
TVA2In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, 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 NRC's 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/
Loyd M. Cain, Acting ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos.:50-390License Nos.:NPF-90 and Construction Permit No. CPPR-92
Enclosure:
NRC Inspection Report 05000390/2007006 w/Attachment: Supplemental Information
REGION IIDocket Nos.:50-390License Nos.:NPF-90 Report Nos.:05000390/2007006 Licensee:Tennessee Valley Authority Facility:Watts Bar Nuclear Plant, Unit 1 Location:Spring City, TN 37381 Dates:February 12 - March 9, 2007 Inspectors:S. Rose, Senior Reactor Inspector (Lead)R. Berryman, Senior Reactor Inspector W. Fowler, Reactor Inspector H. Campbell, Contractor J. Leivo, ContractorApproved by:Loyd M. Cain, Acting Chief, Engineering Branch 1 Division of Reactor Safety 2Enclosure
SUMMARY OF FINDINGS
IR05000390/2007006; 2/12/2007 - 2/16/2007, 2/26/2007 - 3/9/2007; Watts Bar Nuclear PowerPlant; Component Design Bases Inspection.This inspection was conducted by a team of three NRC inspectors and two NRC contractors. One green finding, non-cited violation, was identified during this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609,
"Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The team identified a violation of Technical Specification 5.7.1associated with TVA's failure to develop a procedure that will provide tornado depressurization protection of the emergency diesel generator building. The finding involves a severe weather event (tornado) in which the Emergency Diesel Generator ventilation system would not be properly aligned to prevent inoperability of the Diesel Generators. Abnormal Operating Instruction - 8 (AOI-8) does not provide guidance on how to provide pressure equalization for mitigating atmospheric depressurization associated with tornadic conditions during weather where temperatures are below 68 degrees Fahrenheit. This finding was more than minor because it is associated with the MitigatingSystems Cornerstone attribute of Procedure Quality. It impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The team reviewed the finding for significance using the SDP Phase 1 worksheet for mitigating systems and determined the finding needed to be processed using the SDP Phase 3 due to the external event initiator (tornado).The Regional Senior Reactor Analyst (SRA) performed an SDP Phase 3 for thefinding. The risk of the finding was determined through a hand calculation using various estimates of the significant inputs. There was high uncertainty associated with most of the inputs. Because of this, conservative inputs were used, so the result was more of a bounding analysis than a calculation of the real risk. The dominant risk sequence involved: a tornado occurs onsite when the outside temperature is below 68 degrees, which results in a loss of offsite power; the plant staff fails to recognize and repair the loss of the ventilation system in time to prevent the loss of all the emergency diesels; the resulting loss of all Alternating Current (AC) power causes a Reactor Coolant Pump Seal Loss of Coolant Accident; and AC power is not recovered in time to prevent core damage. The result of the change in Core Damage Frequency calculation was 3EnclosureGreen, primarily due to the low likelihood of the onsite tornado. Because of theamount of time available in the dominant sequence between the degradation in plant conditions and any potential release due to core damage, there would be time available to evacuate areas around the plant prior to any release. Because of this, there is not an increase in the significance of the finding due to Large Early Release Frequency. The finding is
- Green.
B.Licensee-identified Violations None 4
REPORT DETAILS
1.REACTOR SAFETYCornerstones: Mitigating Systems and Barrier Integrity1R21Component Design Bases Inspection (71111.21).1Inspection Sample Selection ProcessThe team selected risk significant components and operator actions for review usinginformation contained in the licensee's Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1 X10E-6. The components selected were located within the emergency raw water cooling (ERCW)system, coolant charging system (CCS), emergency diesel generator (EDG) ventilation, battery, and generator subsystems, and shutdown board (SBDB) ventilation. The sample selection included 19 components, five operator actions, and six operating experience items. Additionally, the team reviewed two modifications by performing activities identified in IP 71111.17, "Permanent Plant Modifications," Section 02.02.a.
and IP 71111.02, "Evaluations of Changes, Tests, or Experiments." The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, Regulatory Issue Summary 05-020 (formerly Generic Letter 91-18) conditions, NRC resident inspector input of problem equipment, system health reports, industry operating experience and licensee problem equipment lists.
Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report..2 Results of Detailed Reviews.2.1Emergency Raw Cooling Water (ERCW) Pumps
a. Inspection Scope
The team reviewed the design basis documentation, including portions of the UpdatedFinal Safety Analysis Report (UFSAR), Technical Specifications, System Descriptionand station drawings to determine the design requirements for the ERCW pumps. The 5team reviewed ERCW system hydraulic and deep well shaft/column thermal expansioncalculations, problem event reports (PERs) and industry operating experience notices.
Further the team reviewed completed surveillances, work orders and preventative maintenance (PMs) for the past several years to assess overall pump operability. In addition, the team walked down portions of the ERCW system to verify that the installed configuration was consistent with design base information and visually inspected the material condition of the pumps, motors and associated piping.
b. Findings
No findings of significance were identified.
a. Inspection Scope
The team reviewed the valve specifications, and those portions of the ventilation systemwiring diagrams and ERCW electrical logic diagrams which described the original and current design function of this flow control valve. It was learned that the valve controls had been electrically disconnected at the Motor Control Center due to potential Appendix 'R' interactions. Discussions with licensee probabilistic safety assessment personnel focused on the impact of the valve's potential failure on core damage frequency (CDF), primarily the concern for valve internal failure via dropped disc or blockage. This led to a review of monthly surveillances performed to ensure a flow path for adequate ERCW coolant to the RHR pump room coolers.
b. Findings
No findings of significance were identified..2.3Centrifugal Charging Pumps (CCPs)
a. Inspection Scope
The CCPs have major Emergency Core Cooling System (ECCS) and charging/RCP-seal-injection functions, both which contribute significantly to the station CDF.
The team reviewed the design basis documentation, including portions of the UFSAR, Technical Specifications, System Descriptions and station drawings to determine the design requirements for the CCPs. Completed surveillances, at both quarterly and 18 month frequencies, work orders and PMs for the past several years for the pumps were reviewed to assess overall pump condition and operability. Discussions with the system and design engineers provided further information on pump performance and requirements. Finally, calculations addressing net positive suction head and vortex concerns were reviewed.
b. Findings
No findings of significance were identified..2.4Chemical Volume and Charging System (CVCS) Charging Header Flow/PressurizerLevel Control Valve
a. Inspection Scope
The team reviewed design documentation as provided in the UFSAR and systemdescriptions, including valve specifications, P&IDs and electrical control diagrams. Further, the team discussed operating history and maintenance with the CVCS system engineer. Finally, the team performed a review of the three most recent performances of the "18 Month Channel Calibration of Charging Header Flow Loop" which addressed the control circuitry and operability of this valve.
b. Findings
No findings of significance were identified..2.5Shutdown Board (SDBD) Air Handling Units (AHUs) Fans
a. Inspection Scope
The team reviewed the design requirements of the SDBD AHUs including portions of theUFSAR, Technical Specifications, flow diagrams, and electrical & control schematics. Heat load calculations and associated AHU flow requirements were reviewed in addition to the flow balance tests and associated test deficiencies and resolutions. Discussions with the system and design engineers were performed. The team performed walkdowns of the fans and motors to assess the system's overall operability and condition.
b. Findings
No findings of significance were identified.
.2.6 SDBD Room Chiller Units
a. Inspection Scope
The team reviewed the design requirements of the SDBD room chillers includingportions of the UFSAR, Technical Specifications, flow diagrams, and electrical & control schematics. Calculations evaluating heat loads during normal and accident conditions in the 480 V and 6.9 KV SDBD rooms were also reviewed. Surveillances of the chiller pumps, associated instrumentation, and PERs addressingchiller trips were discussed with the system engineer. Walkdowns of the chillers, pumps 7and related piping was conducted to ensure conformance with design requirements andto assess the current condition of the system.
b. Findings
No findings of significance were identified..2.7Emergency Diesel Generator (EDG) Building Ventilation Dampers
a. Inspection Scope
The team reviewed the UFSAR, system description, Abnormal Operating Instructions(AOI's), surveillance procedures, control logic, operator actions and calculations to verify the diesel building intake and Electrical Board Room (EBR) ventilation dampers can perform their safety functions of providing cooling to components in the diesel generator building. Automatic controls were compared with operator actions that may be required during certain conditions to determine their feasibility and the ability to detect inadequate ventilation conditions.System walkdowns were performed to verify appropriate missile protection was in place,damper physical condition and linkages could operate as designed, effectiveness of identifying ineffective ventilating conditions and that debris accumulation in ductwork has not adversely affected ventilation flow paths.
b. Findings
Introduction.
The team identified a finding of very low safety significance involving aviolation of Technical Specification 5.7.1 associated with TVA's failure to develop a procedure that will provide tornado depressurization protection of the emergency diesel generator building. AOI-8 does not provide guidance on how to provide pressure equalization for mitigating atmospheric depressurization associated with tornadic conditions during weather where temperatures are below 68 degrees Fahrenheit. The significance of this violation was determined using Phase III of the Significance Determination Process (SDP).Description. The diesel generator building (DGB) ventilation system at Watts Barincorporates a once through cooling method where intake and exhaust dampers are interlocked with the ventilation fans. Upon fan startup both the intake and exhaust dampers receive their individual signals to open and building ventilation will occur by means of a once through cooling method. The only exception to this is if the diesel generator starts. This will then allow the intake damper to receive a signal to open, but if the exhaust room temperatures do not exceed the ventilation fan startup set point building ventilation will not occur, due to the exhaust damper being interlocked with the fan. The UFSAR Section 9.4-33 states that both the DGB ventilation intake and exhaust dampers must be in the open position to mitigate atmospheric depressurization during a 8tornado. AOI-8 provides procedural guidance during a tornado watch and warning foran auxiliary operator to verify that both the intake and exhaust dampers are in the open position. If the dampers are observed to be in the closed position then the ventilation exhaust fans are to be started to provide the needed logic for the dampers to open.Control switches for the exhaust fans are spring return to auto. In the event thatexhaust room air temperature is either initially below the fan shutoff set point of 68 degrees Fahrenheit or in the event room temperatures fall below this during a tornado the ventilation fans will shutoff. Subsequently, both the intake and exhaust dampers will either fail to open or go closed, as the opening logic will no longer be satisfied. This will result in the inability to mitigate atmospheric depressurization of the DGB associated with a tornado. This represents a potential common-cause failure of all four emergency diesel generator ventilation systems and subsequent inoperability of the EDG.Analysis. TVA's failure to develop a procedure that will provide tornado depressurizationprotection of the DGB during temperatures below 68 degrees Fahrenheit is a performance deficiency associated with the Mitigating Systems Cornerstone. Traditional enforcement does not apply because an event did not occur that resulted in an actual safety consequence, the failure to have an adequate procedure did not impact the NRC's regulatory function, and was not the result of a willful violation of NRC requirements or TVA procedures. The finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute Procedure Quality for AOIs. It impacts the cornerstones objective of ensuring the availability, reliability, and operability of the emergency diesel generators to perform their safety function during an initiating event, such as, a loss of offsite power. The finding involved the unavailability of a design feature described in the UFSAR that would ensure tornado depressurization mitigation during atmospheric temperatures below 68 degrees Fahrenheit. The Regional SRA performed an SDP Phase 3 for the finding. The risk of the findingwas determined through a hand calculation using various estimates of the significant inputs. There was high uncertainty associated with most of the inputs. Because of this, conservative inputs were used, so the result was more of a bounding analysis than a calculation of the real risk. The dominant risk sequence involved: a tornado occurs onsite when the outside temperature is below 68 degrees Fahrenheit, which results in a loss of offsite power; the plant staff fails to recognize and repair the loss of the ventilation system in time to prevent the loss of all the emergency diesels; the resulting loss of all Alternating Current (AC) causes a Reactor Coolant Pump Seal Loss of Coolant Accident; and AC power is not recovered in time to prevent core damage. The result of the change in Core Damage Frequency calculation was Green, primarily due to the low likelihood of the onsite tornado. Because of the amount of time available in the dominant sequence between the degradation in plant conditions and any potential release due to core damage, there would be time available to evacuate areas around the plant prior to any release. Because of this, there is not an increase in the significance of the finding due to Large Early Release Frequency. The finding is of very low safety significance (Green).
9Enforcement. Technical Specification 5.7.1, "Procedures", requires in part that writtenprocedures be established, implemented and maintained per Regulatory Guide (RG)1.33, Rev. 2, "Quality Assurance Program Requirements." Appendix A of RG 1.33 states that procedures for combating emergencies and other significant events, such as tornados, shall be covered by written procedures. Contrary to the above, TVA did not develop procedures for mitigating tornado depressurization during temperatures below 68 degrees Fahrenheit. This finding was entered into Watts Bar's Corrective Action Program under PER 120005 and interim actions have been taken to ensure the ability to operate the ventilation exhaust fans at temperature below 68 degrees Fahrenheit until a permanent modification can be implemented. Because the finding is of very low safety significance (Green) and entered into the licensee's corrective action program, this violation is being treated as a non-cited violation (NCV) consistent with section VI.A.1 ofthe NRC Enforcement Policy: NCV 05000390/2007006-01, Violation of Technical Specification 5.7.1 for TVA's failure to develop a procedure that will provide tornado depressurization protection of the DGB.
.2.8 Emergency Diesel Generator Building Generator and Electrical Panel Ventilation Fan
a. Inspection Scope
The team reviewed the UFSAR, system description, Heating Ventilation Air Conditioning(HVAC) duct flow balances and calculations to verify the generator and electrical panel ventilation fans are designed appropriately, such that, they are able to provide sufficient air-flow during elevated temperature conditions to prevent exceeding equipment design temperatures. Fan motor sizing and vane angle position design calculations were reviewed to verify they are adequate for the installed configuration. Also, control logic was reviewed to verify that adequate air supply would be provided for various ambient air temperatures.System walkdowns were performed to identify any potential bypass flow paths, ductworkcondition, debris accumulation and any additional heat loads, which were not taken into account in ventilation analyses.
b. Findings
No findings of significance were identified..2.9CCP 1B-B Room Cooler Fan
a. Inspection Scope
The team reviewed the UFSAR, system description, heat load calculations, surveillanceand maintenance procedures to verify the room cooler fan can provide adequate airflow to maintain room temperatures below design basis temperatures during accident conditions. Flow rates and temperatures used in the auxiliary building heat load nodal analysis were reviewed to identify any discrepancies with the installed configurations 10capabilities. Periodic maintenance and surveillances for the room cooler fan werereviewed to verify they are capable of identify degrading conditions.Systems walkdowns were performed to identify any degrading airflow paths, dissimilarmetal flanges, appropriate valve line-ups, and to verify monitored ERCW flow rates are above the design basis rate at which the fan can sustain the room temperature below the design limits during accident conditions.
b. Findings
No findings of significance were identified..2.10ERCW Piping
a. Inspection Scope
The team reviewed the UFSAR, system description, modification packages, systemlayout drawings, chemistry procedures, wall-thickness monitoring procedures, and PERs to verify the ERCW system will be capable of performing its safety-related functions for providing cooling water to the diesel generators, CCP room coolers and performing the safety-related water supply for the Auxiliary Feedwater system (AFW). Microbiological Induced Corrosion (MIC) monitoring programs were reviewed to verify degradations are being identified and corrected. Licensee discussions were conducted to review the selection criteria of wall-thickness locations and the susceptibility of yard piping to both internal and external corrosion.System walkdowns were performed to verify "A" and "B" ERCW train valve line-upswere correct within the DGB for both the north and south diesels and that indications were consistent on both local controls and EBR breaker panels. Walkdowns were performed on the "A" and "B" ERCW discharge headers to verify the adequacy of float valve installations used to vent header air entrainment and that appropriate equipment was in place to ensure discharge header level was above procedural limits. The walkdowns were performed to ensure that the "B" AFW suction line does not contain air pockets which would prevent the AFW system from performing its' design basis function. The ERCW intake structure was inspected for environmental impacts on the ERCW pumps and motors.
b. Findings
No findings of significance were identified..2.11 ECCS Discharge Relief Valves
a. Inspection Scope
The team reviewed the UFSAR, valve bench testing procedures, and applicableAmerican Society of Mechanical Engineers (ASME) code requirements to verify that 11relief valve springs are sized according to specifications during maintenance activities. Also, the team reviewed functional evaluations associated with inadvertent relief valve lifts and their impact on the safety injection systems ability to provide sufficient flow in the event they do not reseat. The team observed licensee demonstrations of the relief valve bench testing stand andverified personnel follow procedural guidance on action related to valves that do not meet their acceptance criteria.
b. Findings
No findings of significance were identified..2.12Current Limiting Bypass Breaker
a. Inspection Scope
The team reviewed the design and testing of the current limiting reactor bypass breakercontrol circuits to confirm that design basis functional requirements for the control logic were satisfied, and to assess the potential for single failure vulnerabilities and undetected failures. The team also visually inspected external portions of the 480 VAC switchgear for visible material condition and potential vulnerability to hazards or interactions; reviewed the corrective action history for the breakers; and reviewed the results of the most recent preventive maintenance activities for all of the 480 VAC current limiting reactor bypass breakers.
b. Findings
No findings of significance were identified..2.13Generator for the Emergency Diesel Generating Unit
a. Inspection Scope
The team reviewed the EDG 1A-A loading analysis for a design basis event comprisedof a loss of coolant accident, together with a loss of offsite power (LOOP). For selected loads (centrifugal charging pumps, ERCW pumps, component cooling system (CCS)pumps, and containment spray pumps), the team reviewed design input documents, including calculations that determined brake horsepower values, to confirm that worst-case operating modes under accident conditions had been assumed. In addition, the team reviewed recent health reports and corrective action history for the diesel generator, as well as pre-operational loading tests and periodic surveillance tests usedin determining the transient loading capability of the EDG.
b. Findings
No findings of significance were identified.
12.2.14Relays Supporting Closure of EDG Output Breaker to 6.9 kV Buses
a. Inspection Scope
The team reviewed the design and testing of the EDG breaker control circuits to confirmthat design basis functional requirements for the control logic were satisfied, and to assess the potential for single failure vulnerabilities and undetected failures. This included review of permissive circuits, setpoints, and devices for engine speed and output voltage, as well as associated margins established for instrument uncertainty. In addition, the team visually inspected external portions of the 6900 VAC switchgear for visible material condition and potential vulnerability to hazards or interactions; reviewed the corrective action history for the breakers and control devices; and reviewed the results of the most recent preventive maintenance activities for all of the EDG breakers, which included inspection, testing, and overhaul.
b. Findings
No findings of significance were identified..2.15120 Vac Vital Inverters
a. Inspection Scope
The team reviewed the overall vital instrument bus configuration and inverter loadingcalculations, as well as the scope, design criteria, technical evaluation, selected electrical interfaces, and the 10 CFR 50.59 screening for design change notice (DCN)51370A, which had replaced the original equipment inverters. In addition, the team visually inspected external portions of the inverters for visible material condition and potential vulnerability to hazards or interactions; and reviewed the health reports and corrective action history for the inverters.
b. Findings
No findings of significance were identified..2.16Degraded Voltage Protection
a. Inspection Scope
The team selectively reviewed the degraded voltage protection schematics, as well as the calculations that determined the basis for trip and reset setpoints, and associated instrument uncertainties. The team also reviewed the licensee's load flow / voltage drop calculations to determine voltage at selected load terminals under design basis degraded voltage conditions. Selected loads included the containment hydrogen igniters, centrifugal charging pumps, ERCW pumps, component cooling pumps, and containment spray pumps. In addition, the team visually inspected external portions of 13the degraded voltage relays for visible material condition and potential vulnerability tohazards or interactions; and reviewed the testing and corrective action history for the degraded voltage relay circuits and devices.
b. Findings
No findings of significance were identified..2.17Battery for the Emergency Diesel Generating Unit
a. Inspection Scope
The team reviewed the basis for the EDG battery load profile, battery sizing calculations,and voltage drop calculations for selected EDG devices. The team also reviewed the procedures and most recent results for battery surveillances as well as the most recent service and performance test results (the licensee recently replaced all of the cells) and the ground detection instrumentation and ground management procedures. The team reviewed the licensee's history of Direct Current (DC) grounds for the past three years.
In addition, the team performed a visual inspection of the batteries, cells, and connections with respect to material condition, as-found configuration, and vulnerability to potential hazards.
b. Findings
No findings of significance were identified..2.18Reactor Coolant Pump (RCP) Thermal Barrier Cooling Water Return Flow
a. Inspection Scope
Flow rates in the RCP thermal barrier cooling water supply and return lines arecompared, so as to detect a mismatch in flow, which could be indicative of in-leakage to the CCS. The team reviewed the instrument uncertainty and basis for the setpoints and the instrument loops that are designed to alarm and close the thermal barrier cooling water supply and return isolation motor operated valves (MOVs) if a thermal barrier leak was detected. The team also reviewed the loop and schematic diagrams for the isolation MOVs to confirm that design basis functional requirements were satisfied, and that design basis single failure and independence criteria for redundant containment isolation would not be compromised. This included a review of the thermal overload (TOL) heater size and TOL bypass circuit testing for the associated MOVs, to confirm that premature trips would be precluded without unduly compromising motor protection.
The team also visually inspected the flow transmitter installations that were readily accessible, to assess the upstream / downstream process piping and instrument impulse line configuration (i.e. instrument material condition, line slope, and potential for differential heating). In addition, the team reviewed a sample of calibration / test procedures and results for these instrument loops, and assessed the corrective action history.
b. Findings
No findings of significance were identified..2.19Motors and Control Circuits for ERCW, CCP, and Fans
a. Inspection Scope
For these loads, the team reviewed the adequacy of motor terminal voltage underdegraded voltage conditions. The team also reviewed the design and testability of the breaker control circuits to confirm that functional requirements for the design basis control logic were satisfied, and to assess the potential for single failure vulnerabilities and undetected failures. In addition, the team reviewed calculations that established the basis for protective trip settings and tolerances, to confirm that premature trips would be precluded without unduly compromising motor and feeder protection. In conjunction with an assessment of the licensee's disposition of NRC IN 2002-12,Submerged Safety-Related Cable (listed in Section 1R21.4a of this report), for selected ERCW pump motor feeder cables, the team reviewed a sample of insulation resistance test results and manhole inspections associated with underground cable with potential for submergence. The underground cables connecting the EDGs to the associated 6900 VAC switchgear buses were also included in this sample.
b. Findings
No findings of significance were identified..3Review of Low Margin Operator Actions
a. Inspection Scope
The team performed a margin assessment and detailed review of five risk significantand time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures (JPMs). For the selected components and operator actions, the team performed an assessment of the Emergency Operating Instructions (EOIs), AOIs, Alarm Response Instructions (ARIs), and other operations procedures to determine the adequacy of the procedures and availability of equipment required to complete the actions. Operator actions were observed on the plant simulator and during plant walkdowns. The following operator actions were observed on the licensee's operator trainingsimulator:*Operation of the containment hydrogen igniters: E-1, Loss of Reactor orSecondary Coolant 15*Actions to reestablish CCS flow to RCP thermal barriers after phase-Bcontainment isolation signal: ARI-125-B, Containment Hi-Hi Pressure Spray Actuate*Loss of ERCW: AOI-13 Loss of Essential Raw Cooling Water
- Operation of TDAFW pump and establishing flow after station blackout: ECA0.0, Loss of Shutdown Power, AOI-40, Station Blackout, and SOI-3.02, Auxiliary Feedwater SystemAdditionally, the inspectors walked down, "table-topped" and investigated the followingoperational scenarios:*Manual actions to install spool pieces to supply CCS from ERCW duringflooding event: MI-17.021, Installation of Spool Pieces Between ERCW System and Component Cooling System*Local Operation of TDAFW Pump without electric instrumentation: SOI-3.02,Auxiliary Feedwater Systemb.FindingsNo findings of significance were identified.
.4 Review of Industry Operating Experience
a. Inspection Scope
The team reviewed selected operating experience issues that had occurred atdomestic and foreign nuclear facilities for applicability at the Watts Bar Nuclear Plant.
The team performed an independent applicability review, issues that were identified as applicable to the Watts Bar Nuclear Power Plant were selected for a detailed review.
The issues that received a detailed review by the team included:*NRC Event Notification 42844 "Improper Valve Springs in Pressure Relief Valves"*NRC Information Notice (IN) 94-45, Potential Common Mode Failure Mechanismsfor Large Vertical Pumps
b. Findings
No findings of significance were identified..5Review of Permanent Plant Modifications
a. Inspection Scope
The team reviewed two modifications related to the selected risk significantcomponents in detail to verify that the design bases, licensing bases, and performance capability of the components have not been degraded through modifications. The adequacy of design and post modification testing of these modifications was reviewed by performing activities identified in IP 71111.17, "Permanent Plant Modifications,"
Section 02.02.a. Additionally, the team reviewed the modifications in accordance IP 71111.02, "Evaluations of Changes, Tests, or Experiments," to verify the licensee had appropriately evaluated them for 10 CFR 50.59 applicability. The following modifications were reviewed:*DCN 52011, Install Pressure Air Release Valves on ERCW Discharge Headers "A"& "B"*DCN-52179, Shutdown Board Room A and B Air Handling Unit A-A MotorReplacement and Circuit Protective Devices Settings Changes, Rev. A
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4AO6Meetings, Including ExitExit Meeting SummaryOn March 9, 2007, the team presented the inspection results to Mr. M. Lorek, PlantManager, and other members of the licensee staff. The team returned all proprietary information examined to the licensee. No proprietary information is documented in the report.On April 18, 2007, a telephone exit was conducted to present the results of the SDPPhase 3 for the DGB ventilation finding (Section 1R21.2.7) to Mr. M. Lorek and other members of the licensee staff.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Lorek, Plant Manager
- W. Justice, General Manager TVAN Engineering
- T. Carter, Manager, Engineering Design
- B. Thomas, Site Licensing
- J. Smith, Site Licensing and Incident Response Manager
- C. Borrelli, PSA Specialist
NRC
- J. Bartley, Senior Resident Inspector, Watts Bar Nuclear Plant
- C. Ogle, RII, Engineering Branch 1, Chief
ITEMS OPENED, CLOSED, AND DISCUSSED
Open/Closed05000390/2007006-01NCVViolation of Technical Specification 5.7.1 for TVA's failure todevelop a procedure that will provide tornado
depressurization protection of the DGB. (Section 1R21.2.7)
2Attachment
DOCUMENTS REVIEWED
Updated Final Safety Analysis ReportUFSAR Section 9.4, Air Conditioning, Heating, Cooling, and Ventilation SystemsUFSAR Section 9.4.5.2.1, Diesel Generator Building