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MONTHYEARML0630602512003-10-15015 October 2003 WB1-DWD-015A, Containment Debris Walkdown Package Watts Bar N. P. Unit 1. Project stage: Other ML0629105222003-11-0505 November 2003 WB1-DWD-024A, Containment Debris Walkdown Package Watts Bar N. P. Unit 1. Project stage: Other ML0629105132004-08-24024 August 2004 TVAW001-RPT-001, Rev. 0, Report on Watts Bar, Unit 1 Containment Building Walkdowns for Emergency Sump Strainer Issues. Project stage: Other ML0515205802005-06-0303 June 2005 Request for Additional Information Related to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-water Reactors Project stage: RAI ML0603704802006-02-10010 February 2006 Request for Additional Information Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design-basis Accidents at Pressurized-water Reactors Project stage: RAI ML0603800822006-02-10010 February 2006 RAI, Response to GL-2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized-water Reactors Project stage: RAI ML0610203132006-04-11011 April 2006 Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized Water Reactors - Response to Request for Additional Information Project stage: Response to RAI ML0610402192006-04-11011 April 2006 Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized Water Reactors - Response to Request for Additional Information Project stage: Response to RAI ML0610101802006-05-10010 May 2006 Request for Additional Info the Nuclear Regulatory Commission Staff Audit on the Containment Sump Modification Project stage: Acceptance Review ML0621204722006-07-0303 July 2006 Watts Bar Nuclear Plant (WBN) Unit 1 - Generic Letter 2004-02 - Request for Additional Information Regarding the Nuclear Regulatory Commission Staff Audit on the Containment Sump Modifications Project stage: Request ML0629105082006-07-0303 July 2006 Generic Letter 2004-02 - Request for Additional Information Regarding NRC Staff Audit on Containment Sump Modifications Project stage: Request ML0621800772006-08-11011 August 2006 Proprietary Letter, Request for Withholding Information from Public Disclosure, Containment Sump Modifications Project stage: Withholding Request Acceptance ML0636204112006-12-21021 December 2006 Nuclear Regulatory Commission (NRC) Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Supplemental Response Project stage: Supplement ML0721501092007-08-0101 August 2007 Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized Water Reactors - Request for Extension of Completion Date for Corrective Actions Project stage: Request ML0727604492007-10-0202 October 2007 Watts Bar Nuclear Plant (WBN) Unit 1, Generic Letter 2004-02, Recirculation During Design-Basis Accidents at Pressurized Water Reactors - Request for Extension of Completion Date for Corrective Actions Project stage: Request ML0733703172007-11-28028 November 2007 Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (PWR) - Notice of Completion Project stage: Other ML0733801522007-12-0606 December 2007 Generic Letter 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors Extension Request Evaluation Project stage: Other ML0806402052008-02-29029 February 2008 Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accident at Pressurized-Water Reactors (PWR) - Notice of Completion Project stage: Supplement ML0811201592008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Calculation No. PCI-5343-S01, Rev. 0, Attachment D, Page 374 of 575 Through Calculation No. PCI-5343-S02, Rev 0, Attachment B, Page 100 of 631 Project stage: Supplement ML0810905032008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Calculation No. PCI-5464-S01, Rev. 2, Attachments a and B, Page 384 of 680 Project stage: Supplement ML0810905022008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Figure 4.25 - Case 3 Through Calculation PCI-5464-S01, Page 107 Project stage: Supplement ML0810905002008-03-31031 March 2008 Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors - Notice of Completion Project stage: Supplement ML0810905012008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Pages E1-1 Through Figure 4.24 - Case 3 Project stage: Supplement ML0810905082008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Calculation No. PCI-5464-S02, Rev. 0, Attachment B Page 101 Through End Project stage: Supplement ML0810905052008-03-31031 March 2008 Enclosure 1, Supplemental Response Addressing GL-04-002 Actions, Calculation No. PCI-5464-S01, Rev. 2, Attachment B, Page 385 Through 680 Project stage: Supplement ML0832308232008-11-25025 November 2008 Request for Additional Information Regarding Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors (Tac Nos. MC4717 and MC4718) Project stage: RAI ML0833700332008-12-0303 December 2008 Request for Additional Information Regarding Generic Letter 2004-02. Potential Impact of Debris Blockage During Design Basis Accidents at Pressurized-Water Reactors (Tac No. MC4730) Project stage: RAI ML0905408572009-02-23023 February 2009 Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized Water Reactors - Response to Request for Additional Information Project stage: Response to RAI ML0907208692009-02-25025 February 2009 Calculation, ALION-CAL-TVA-2739-03, Rev. 4, Watts Bar Reactor Building GSI-191 Debris Generation Calculation. Project stage: Other ML0907208682009-03-0303 March 2009 Response to Request for Additional Information Regarding Generic Letter 2004-02,.Potential Impact of Debris Blockage During Design Basis Accidents at Pressurized-Water Reactors. Project stage: Response to RAI ML0926502602009-09-29029 September 2009 RAI, Regarding Generic Letter 2004-02, Potential Impact of Debris Blockage During Design Basis Accidents at Pressurized-Water Reactors Project stage: RAI ML0936502182009-12-23023 December 2009 GL 2004-02 Final RAI Responses Due Date Project stage: RAI ML1000601162010-01-0707 January 2010 Change of Response Date Regarding Request for Additional Information for Generic Letter 2004-02 Potential Impact of Debris Blockage During Design-Basis Accidents at Pressurized Project stage: RAI ML1015905562010-06-0909 June 2010 Request for Withholding Proprietary Information from Public Disclosure Related to Generic Letter 2004-02 Project stage: Other ML1016600442010-06-17017 June 2010 Summary of Conference Call with TVA on Proposed Response to Request for Additional Information on Generic Letter 2004-02 Project stage: RAI ML1107000312011-04-15015 April 2011 Notice of Meeting with Tennessee Valley Authority to Discuss Responses to NRC Requests for Additional Information for Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accident at Pre Project stage: RAI ML11152A1702011-05-12012 May 2011 05/12/2011-Meeting Slides from Meeting with TVA on Their Proposed Response to a Request for Additional Information on Generic Letter 2004-02 Project stage: Response to RAI ML11154A0932011-06-28028 June 2011 Request for Withholding Proprietary Information from Public Disclosure Related to Generic Letter 2004-02 Project stage: Other ML11152A1632011-06-28028 June 2011 Summary of Meeting with TVA on Their Proposed Response to a Request for Additional Information on Generic Letter 2004-02 Project stage: RAI ML11230B2502011-11-0303 November 2011 Request for Withholding Proprietary Information from Public Disclosure Related to Generic Letter 2004-02 Project stage: Other CNL-14-114, Resolution of Generic Safety Issue (GSI)-1912014-06-27027 June 2014 Resolution of Generic Safety Issue (GSI)-191 Project stage: Request ML14283A5132014-11-17017 November 2014 NRC Staff Review Documentation Provided by TVA for the Sequoyah Nuclear Plant, Units 1 and 2 Concerning Resolution of GL2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurizer-Wate Project stage: Approval ML14283A5262014-11-17017 November 2014 Closeout of GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactor Project stage: Other CNL-15-009, Resolution of Generic Safety Issue 1912015-04-17017 April 2015 Resolution of Generic Safety Issue 191 Project stage: Request ML15191A1832015-09-17017 September 2015 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: Other CNL-15-252, License Amendment Request and Request for Deviation from 10 CFR 50, Appendix R in Support of Closure of Generic Safety Issue 191 (WBN-TS-16-01)2016-02-23023 February 2016 License Amendment Request and Request for Deviation from 10 CFR 50, Appendix R in Support of Closure of Generic Safety Issue 191 (WBN-TS-16-01) Project stage: Request 2008-03-31
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Category:Slides and Viewgraphs
MONTHYEARML24037A0252024-02-0808 February 2024 Pre-Submittal Presentation - 02/08/2024 - Proposed License Amendment Request Change to the Sequoyah Nuclear Plant Technical Specifications 3.8.1 and 3.8.2 ML23345A1662023-12-13013 December 2023 Tennessee Valley Authority - Pre-submittal Meeting for Conversion of Nuclear Quality Assurance Plan from ANSI N45.2-1971 to ASME NQA-1-2015 (EPID L-2023-LRM-0103) - Slides ML23291A1562023-10-18018 October 2023 NRC Pre-Submittal Meeting Watts Bar & Sequoyah LAR, TVA Presentation Slides ML23264A0412023-09-27027 September 2023 NRC Pre-Submittal Meeting WBN Rebaseline LAR, TVA Presentation Slides ML23251A1252023-09-12012 September 2023 Pre-submittal Meeting for License Amendment Request, Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specification Surveillance Requirement 3.9.5.1 ML23251A1212023-09-12012 September 2023 RHR Flow Rate Reduction During Mode 6 Operation at 23 Ft. Refueling Water Level Technical Specification Change NON-PROPRIETARY Slides ML23178A0832023-07-10010 July 2023 Pre-submittal Meeting Slides for License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs ML23128A2882023-05-0808 May 2023 TVA Nuclear Plants -2022 EOC Media Slides ML23052A0362023-02-22022 February 2023 February 22, 2023, Pre-Submittal Meeting for the License Amendment Request for Increased Tritium Production at Watts Bar (EPID L-2023-LRM-0010) - Meeting Slides ML23044A5182023-02-15015 February 2023 Pre-Submittal Meeting Slides for Expedited License Amendment Request Re Main Control Room Chiller Completion Time Extension ML23041A3562023-02-15015 February 2023 February 15, 2023 Pre-submittal Meeting for Expedited License Amendment Request Main Control Room (MCR) Chiller Completion Time Extension (EPID L-2022-LRM-0098) (Slides) ML23010A2542023-01-12012 January 2023 Pre-submittal Meeting for Expedited License Amendment Request Main Control Room (MCR) Chiller Completion Time Extension -- Meeting Slides ML22318A1042022-11-14014 November 2022 Pre-submittal Meeting for American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative WBN-2-ISI-01 Slide Presentation ML22201A1262022-07-20020 July 2022 July 20, 2022 - Pre-submittal Meeting for Inservice Testing (IST) Program Request for Alternative for the Centrifugal Charging Pump 1B-B Testing Per ISTB-3310 (RP-12) ML22157A4132022-06-0606 June 2022 Plants - Pre-Submittal Teleconference for Proposed Amendment Request Regarding the Change for the Sequoyah and Watts Bar Nuclear Plants Technical Specification 3.4.12 Meeting Slides ML22103A0852022-04-14014 April 2022 and Watts Bar Plant Pre-submittal Meeting for Inservice Testing (IST) Program Request for Alternative for the Motor Driven Auxiliary Feedwater Pump Testing Following Maintenance During an Outage Per ISTB-3310 ML22102A3032022-04-12012 April 2022 Pre-submittal Meeting for Request for Exemption from Requirements of 10 CFR 26.205(d)(4) and 10 CFR 26.205(d)(6) Slide Presentation ML22080A1662022-03-21021 March 2022 TVA Nuclear Plants -2021 EOC Media Loop Slides for Automatic Play ML22046A2932022-02-15015 February 2022 Pre-submittal Meeting Presentation - Request for Alternative to ASME OM Code ISTC-3630 Regarding Leakage of Pressure Isolation Valves ML22035A0002022-02-0404 February 2022 February 8, 2022, Pre-Submittal Meeting Slides Regarding Proposed Change to Modify the Allowable Value for Technical Specification Table 3.3.2-1, Function 6.e(1) ML22013A6432022-01-19019 January 2022 TVA Slides - Watts Bar Nuclear Plant (WBN) Schedule for the Development of a License Amendment Request (LAR) for Increased Tritium Production at WBN ML21343A0282021-12-14014 December 2021 TVA Slide Presentation Regarding Pre-Submittal Meeting for Proposed Revision to Capsule Withdrawal Schedule ML21308A4932021-11-0808 November 2021 Pre-submittal Meeting Presentation - License Amendment Request to Revise Sequoyah Operating License to Use Fire and Seismic PRA Models for 10 CFR 50.69 Risk-Informed Categorizations (EPID L-2021-LRM-0109) (Slides) ML21281A2282021-10-13013 October 2021 Tennessee Valley Authority Fleet Pre-Submittal Meeting for License Amendment Request to Change Emergency Action Level Scheme Presentation - October 13, 2021 (EPID L-2021-LRM-0096) (Slides) ML21252A3632021-09-15015 September 2021 Pre-Submittal Meeting for License Amendment Request to Eliminate the High Negative Flux Rate Trip ML21250A3652021-09-0808 September 2021 Meeting Slides - Watts Bar Nuclear Plant Unit 2, Presubmittal Teleconference for Proposed License Amendment to Revise the Updated Final Safety Analyses Report (UFSAR) to Use a Tube Support Plate Locking/Displacement Analysis ML21217A2872021-08-19019 August 2021 TVA Slides - Pre-Submittal Meeting for Proposed License Amendment Request Re Application to Modify Technical Specification 3.7.8 Essential Raw Cooling Water (ERCW) System, to Support Permanent Shutdown Board ML21229A1992021-08-19019 August 2021 Mid-Cycle SG Outage Strategy (EPID L-2021-LRM-0067) (Slides) ML21231A0972021-08-18018 August 2021 TVA Slides - Pre-Submittal Meeting for Proposed LAR Regarding Application to Modify Watts Bar Units 1 and 2 Technical Specification 3.7.8, Essential Raw Cooling Water (ERCW) System, to Support Permanent Shutdown Board Cleaning ML21197A0022021-08-0505 August 2021 N-769-6 Relief Pre-Submittal TVA Presentation August 2021 ML21111A0202021-04-21021 April 2021 TVA Slides for April 21, 2021 Public Meeting to Discuss Planned Sequoyah Amendment Request ML21096A1312021-04-0707 April 2021 Pre-submittal Meeting for Inservice Testing (IST) Program Request for One-Time Extension of Test Frequency of WBN Unit 1 Group 6 Relief Valves ML21096A1002021-04-0606 April 2021 Annual Assessments ML21061A0892021-03-0202 March 2021 TVA Presentation Slides for March 2, 2021, Public Meeting to Discuss Sequoyah Alternative Request ML21041A1552021-02-11011 February 2021 Meeting Slides for February 11, 2021 Meeting with TVA ML21019A5692021-01-19019 January 2021 Pre-Submittal Teleconference for Proposed License Amendment Requests (Lars) for the WBN Unit 2 Replacement Steam Generator (RSG) Project ML20352A0362020-12-17017 December 2020 Pre-Submittal Teleconference for Proposed License Amendment to Revise the Updated Final Safety Analysis Report (UFSAR) to Use an Alternate Probability of Detection (POD) ML20329A3722020-11-24024 November 2020 Pre-Submittal Teleconference for Proposed License Amendment Request Regarding Technical Specification 5.9.9, Steam Generator Tube Inspection Report, to Use Popcd Methodology ML21050A1562020-10-21021 October 2020 Graizer - Presentation - Analysis of Tva'S Watts Bar Nuclear Power Plant Strong-Motion Records of the M 4.4 December 12, 2018 Decatur Tennessee Earthquake ML20239A8672020-08-27027 August 2020 Pre-Submittal Meeting Slides License Amendment Request for Technical Specification 5.7.2.19 ML20220A2022020-08-0707 August 2020 Pre-Submittal Meeting Slides - Proposed License Approach Regarding a Change for Watts Bar Nuclear Plant Units 1 & 2 Technical Specification 3.3.2 Table 1 Function 6.e Auxiliary Feedwater Auto-Start from Loss of Main Feedwater Pumps(Epid L-2 ML20209A0702020-07-27027 July 2020 Pre-submittal Meeting for Inservice Testing (IST) Program Request for Relief from Increased Frequency Testing of Residual Heat Removal (RHR) Pump 1B-B Meeting Slides ML20196L6892020-07-16016 July 2020 TVA Non-Proprietary Slides for Open Session of July 16, 2020, Partially Closed Public Meeting to Discuss Sequoyah License Amendment Request ML20203M0572020-07-16016 July 2020 Updated TVA Non-Proprietary Slides for Open Session of July 16, 2020, Partially Closed Public Meeting to Discuss Sequoyah License Amendment Request ML20196L7922020-07-16016 July 2020 Corrected TVA Non-Proprietary Slides for Open Session of July 16, 2020, Partially Closed Public Meeting to Discuss Sequoyah License Amendment Request ML20169A5222020-06-25025 June 2020 Pre-Submittal Meeting for License Amendment Request (LAR) Revised Steam Generator (SG) Inspection Intervals and TSTF-510 ML20175A0552020-06-23023 June 2020 TVA Slides for June 23, 2020, Emergency Plan Change Presubmittal Meeting ML20163A6952020-06-12012 June 2020 Bpg Presentation on TVA SCWE Final 2 ML20148M1452020-05-27027 May 2020 (Wbn)Proposed License Approach Regarding a Change for the WBN Units 1& 2 Technical Specification (TS) 3.3.2 Table 1 Function 6.e Auxiliary Feedwater Auto-Start from Loss of Main Feedwater Pumps ML20147A0132020-05-26026 May 2020 NRC Slides Final WBN-1 TS 3.3.3 LAR 2024-02-08
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TENNESSEE VALLEY AUTHORITYSEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2WATTSBARNUCLEARPLANTUNIT1 WATTS BAR NUCLEAR PLANT , UNIT 1NRCMeeting NRC MeetingRegarding Generic Letter 2004-02 Responses Rockville, MarylandMay 12, 2011 Agenda*Introduction
- Review of TVA Small Break Loss-of-Coolant Rod KrichAccident Water Level Calculations
-Sequoyah Nuclear Plant (SQN)
Sump Strainer Submergence Chris Carey
-Watts Bar Nuclear Plant (WBN)
Sump Strainer Submergence
- ReviewofWBNSumpStrainerStructural Robert Kirkpatrick Robert Kirk p atrick Review of WBN Sump Strainer Structural Integrity Calculation
- Applicability of Sump Strainer Structural IntegrityCalculationtoSQN/Plansfor p Dave Lafever Integrity Calculation to SQN/Plans for Insulation Remediation
- Schedule for Final Request for Additional InformationResponseSubmittal Kara Stacy Information Response Submittal*Closing Remarks Rod Krich 2
SQN Sump Strainer Submergence
Background
- NRC Requested TVA Provide Additional Information for SQN
-Demonstrate Adequate Sump Performance during Small BreakLoss-of-Coolant Accident (SBLOCA)
- NRC Concern is Tall Sump Strainers will be Partially Submerged when Engineered Safety Features (ESF) Pumps begin to take Suction from
Sump *Condition could Result in Less Water Flowing through Strainers than being Drawn by ESF Pumps and Cause Loss of Emergency Core CoolingSystem(ECCS)and/orContainmentSpray(CS)
Cooling System (ECCS) and/or Containment Spray (CS)3 SQN Sump Strainer Submergence (continued)ESF Design
- ECCS and CS Pum p Suctions Initiall y Ali gned to Refuelin g Wate rpyggStorage Tank (RWST)
- ECCS Pumps Aligned to Sump on RWST Low Level and CS Pumps AlidtSRWSTLLLl Ali gne d t o S ump on RWST L ow-L ow L eve l*CS Actuates on High-High Containment Pressure (2 psig)
- Containment Air Return Fans Force Steam and Hot Air in Lower Compartment through Ice Condenser
-Steam is condensed
-Hot Air is cooled
-Melt Water from Ice is produced
- Containment Desi gn Channels All Water from CS and Steam 4 gCondensation/Ice Melt Back to Sump SQN Sump Strainer Submergence (continued)SBLOCA Characteristics
- Break Flow Small enough that High Head ECCS Pumps Maintain Pressurizer Level*ECCS Pumps Provide Additional Water to RCS to Maintain Constant RCS Water Volume as RCS Cools
-Requirement Identified by NRC in Discussions with TVA/WBN, Unit 1
- RCS Pressure Remains above Accumulator PressurePreentsAccmlatorandLoHeadSafetInjection
-Pre v ents Acc u m u lator and Lo w Head Safet y Injection*Break Could be Located Such that Liquid Portion of Break Flow is Contained Inside Reactor Cavity (does not Fill Sump)
-This Break Location is not a Design Basis Accident LocationNot in Hot or Cold Leg Pipe
-This Break would not Generate Debris that could be Transported to Sump 5*CS Actuates for Even Smallest Breaks due to Buildup of Steam and Hot Air in Lower Compartment of Containment SQN Sump Strainer Submergence (continued)Original SQN SBLOCA Sump Level Calculation
- UsedOnlyPortionofWaterAvailableinRWSTbetweenRWST Used Only Portion of Water Available in RWST between RWST Minimum Full Level and RWST Low Level
- AssumesReactorCavityFilledpriortoRWSTLowLevel Assumes Reactor Cavity Filled prior to RWST Low Level-Only Occurs if CS does not Actuate
- DidnotAccountforContributiontoSumpWaterVolumefromSteam
- Did not Account for Contribution to Sump Water Volume from Steam Condensing and Ice Melting 6
SQN Sump Strainer Submergence (continued)Revised SQN SBLOCA Sump Level Calculation
- AccratelDeterminesWaterAailableinRWSTbeteenRWST
- Acc u ratel y Determines Water A v ailable in RWST bet w een RWST Minimum Full and RWST Low Level
-Still includes Adverse Instrument Errors
-Change increases Water Volume in Containment by 35,000 gallons
- Accounts for Time Dependent Filling of Reactor Cavity, based on SizefSBLOCA o f SBLOCA-Change Results in Less Water in Reactor Cavity at RWST Low Leveland More Water in Sump 7
SQN Sump Strainer Submergence (continued)Revised SQN SBLOCA Sump Level CalculationClltiDtidLBdAtfIthtldb
- C a l cu l a ti on D e t erm i ne d L ower B oun d on A moun t o f I ce th a t wou ld b e Melted by Steam from Break
-Am ou n t o f Stea m R e l eased fr o m Br ea k based o n Satu r ated Co n d i t i o n s at outoSteaeeasedoeabasedoSatuatedCodtosatRCS Pressure of 600 psia, Accumulator Pressure
-Released Steam Assumed to Condense on Ice and Flow to Sump as SaturatedWateratContainmentPressureof164 psia Saturated Water at Containment Pressure of 16.4 psiaLowest Containment Pressure that CS would be in Operation
-M e l t W ate r fr o m I ce a l so A ssu m ed to Fl o w to Su m p as Satu r ated W ate r at etateoceasossuedtootoSupasSatuatedateatContainment Pressure of 16.4 psiaMinimizes Amount of Ice that is Melted
-Ice Melt Predicted by this Method is Low Compared to Better Estimate Modeling of Ice Condenser by TVA's version of CONTEMPT 8
SQN Sump Strainer Submergence (continued)Revised SQN SBLOCA Sump Level Calculation
- ExplicitlyEvaluatedfollowingBreaksinsideReactorCavity
- Explicitly Evaluated following Breaks inside Reactor Cavity-100 gpm -Just above Break Size that would be Considered LOCA (i.e., Greater than Normal Makeup Capability)
-1200 gpm -Bounding Flow Rate for One-Train of ECCS when RCS remains Pressurized above Accumulator Pressure2500gpmBoundingFlowRateforTwoTrainsofECCSwhenRCSremains
-2500 gpm -Bounding Flow Rate for Two-Trains of ECCS when RCS remains Pressurized above Accumulator Pressure
- Breaks Larger than 2500 gpm cannot be Maintained above Accumulator Injection Pressure
-Accumulator Injection increases Water Volume in Containment 9
SQN Sump Strainer Submergence (continued)Insights from Revised SQN SBLOCA Sump Level Calculation
- About112gallonofMeltWaterFlowstoSumpforeverygallonremoved
- About 1.12 gallon of Melt Water Flows to Sump for every gallon removed from RWST by ECCS
- One CS Pum p in O p eration Results in Hi gher Sum p Level than Two CS ppgp Pumps in Operation
-Two CS Pumps increase Holdup of RWST Water in Refueling Canal
-More Ice Melt due to Longer Time required to Drain RWST to Low Level 10 SQN Sump Strainer Submergence (continued)Revised SQN SBLOCA Sump Level Calculation Results
- SumpWaterLevelatRWSTLowLevel Sump Water Level at RWST Low Level-ECCS Pumps Aligned to Sump
- WhenTwoCSpumpsareinOperation
- When Two CS pumps are in Operation-100 gpm -5.73 feet (4.36 feet of strainer submerged)
-1200 gpm-5.83 feet (4.45 feet of strainer submerged)
-2500 gpm-594feet(457feetofstrainersubmerged) 2500 gpm-5.94 feet (4.57 feet of strainer submerged)
- Above Sump Water Level are Greater than Previously Determined SBLOCA Sum p Level of 2.5 feet at RWST Low Level p-Increase due toCrediting Additional Water Volume in RWSTAccounting for Melt Water Addition to SumpAddiSTiDdtClltihihdAddi ng S ome Ti me D epen d ency t o C a l cu l a tion, w hi c h re d uces Water Holdup in Reactor Cavity 11 SQN Sump Strainer Submergence (continued)Revised SQN SBLOCA Sump Level Calculation Results
- FullSubmergenceofTallStrainersOccurs<47minutesafterRWST Full Submergence of Tall Strainers Occurs < 4.7 minutes after RWSTLow Level due to Continued Operation of CS Pumps with their Suctions Aligned to RWST
-Due to Low Sump Flow Rate (<
2,500 gpm) and Short Time Period, no Significant Debris Accumulation Occurs prior to Full Submergence
-Debris Load for SBLOCA is < 25 p ercent of Debris Load for LBLOC A pNo Potential for Significantly Loading Strainers
-When CS Pumps are aligned to Sump (at RWST Low-Low Level), Sump Water Level is 8.5 feet or 1 foot above top of Tall Strainers (sump flow
> 5,100 gpm and < 12,500 gpm)
- SumpPerformanceforSBLOCAis Acceptable Sump Performance for SBLOCA is Acceptable 12 WBN, Unit 1, Sump Strainer Submergence Overview*AllWaterLevelswillFullySubmergeStrainer
- All Water Levels will Fully Submerge Strainer*Minimum SBLOCA Water Level increased to 5.78 feet from 5.48 feet
- SBLOCA Water Level increases over Time
-With Minimum Level at ECCS Switchover
- Hold Up in RCS due to decreasing Temperature Accounted for in Calculation 13 WBN, Unit 1, Sump Strainer Submergence (continued)
Assumptions
- AllAssumptionsareConservativeandTakentoMinimizeSumpWater All Assumptions are Conservative and Taken to Minimize Sump Water Level*WBN , Unit 1 , Water Level Assum p tions Consistent with SQN Assum p tions,,pp-Minimum Injection from RWST
-No Accumulator Injection
-Reactor Coolant System (RCS) Break Location between Reactor Vessel and Biological Shield Wall
-Maximum RCS Makeu p due to Fluid Shrinka g e pg-Minimum Ice Melt 14 WBN, Unit 1, Sump Strainer Submergence (continued)
Assumptions
- MinimumInjectionfromRWST Minimum Injection from RWST-Fluid Assumed at Maximum Temperature of 105°F to Minimize Water Mass in Tank
-Beginning Water Level at Minimum Operating Level
-ECCS and CS Switchovers to Sump at Low and Low-Low Water Level Setpoint at Upper Analytical Limits
-No Fluid from Accumulators Credited in SBLOCA Water Level Calculation
- RCS Break Location between Reactor Vessel and Biological Shield Wall
-Break Location Chosen so all ECCS Injection will be held up in Reactor Cavity 15 WBN, Unit 1, Sump Strainer Submergence (continued)
Assumptions
- MaximumRCSMakeupduetoFluidShrinkage Maximum RCS Makeup due to Fluid Shrinkage-RCS Pressure and Temperature Assumed to Decrease at (Bounding)
Rates Defined by 2-inch Line Break
-Long-Term Shrinkage Limited to Accumulator Check Valve Pressure of
600 psia*Minimum Ice Melt
-Ice Melt Previously Only included up to RHR Switchover
-Ice Melt Now included until CS Switchover to Sump
-Ice Melt Water and Vapor Exit Temperatures Conservatively taken to
minimiz e I ce M e l t a n d in c r ease Se n s i b l e H eat A dd i t i o neceetadceaseSesbeeatddto 16 WBN, Unit 1, Sump Strainer Submergence (continued)
Analysis*MethodologynotChangedbetweenRevisions
- Methodology not Changed between Revisions *RCS Holdup due to Fluid Shrinkage now Explicitly Calculated
- Calculation Accounts for Holdup Quantities in RCS, ECCS, and CS Piping, Containment Atmosphere, and Physical Locations in
Containment
- Fluid Volume Contributed from RCS Break and Ice Melt is Calculated using WBN Containment Dynamic Response Model based on WCAP 8282 WCAP-8282*Sump Water Volume then Calculated by Interpolation using Fluid VolumefromModel 17 Volume from Model WBN, Unit 1, Sump Strainer Submergence (continued)
Summary*WBNUnit1SumpStrainerwillRemainSubmergedforAllSBLOCA
- WBN , Unit 1 , Sump Strainer will Remain Submerged for All SBLOCA Accident Cases
- RCSShrinkageNowExplicitlyCalculated RCS Shrinkage Now Explicitly Calculated
- Assumptions are Conservative and Documented in Calculation 18 WBN, Unit 1, Sump Strainer Structural Integrity Overview*InitialDesignBasisDebrisLoadedThinBedTestPerformedatAlden Initial Design Basis Debris Loaded Thin Bed Test Performed at Alden Laboratories, July 2010, resulted in Unacceptably High Head Loss
- Further Testin g Reduced Debris Head Loss g-But only by Removing Min-K Insulation
- As-TestedHeadLosshadNegativeImpactonStrainerAssembly As Tested Head Loss had Negative Impact on Strainer Assembly Structural Integrity
- Strainer Assembl y Structural Re quirement Bounds ECCS and CS Pum p yqpNet Positive Suction Head (NPSH) Requirements 19 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
Overview*ReducingHeadLossandAcceptanceofWBNUnit1StrainerTesting Reducing Head Loss and Acceptance of WBN , Unit 1 , Strainer Testing Results Requires Additional Analysis and Modifications
- Three Approaches Considered for Addressing Test Results 1.Develop New Clean Strainer Head Loss (CSHL) Calculation using Computational Fluid Dynamics (CFD) Modeling of Strainer to Reduce CSHL 2.Determine Maximum Structural Qualification of Strainer Assembly 3.Modification to Reduce Debris Head Loss
- Pursuing Three Approaches Led to Following Solutions to increasedHeadLoss Head Loss 1.Replacement of Plenum Cover Plate with Larger Orifice Sizes to Reduce CSHL based on Results of New Calculation 2MiSttlQlifitifStiAbl 2.M ax i mum St ruc t ura l Q ua lifi ca ti on o f St ra i ner A ssem bl y 3.Removal of Min-K Insulation to Reduce Debris Head Loss 20 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
CSHL Reduction
- CFDAnalysisPerformedtoEvaluateCSHL CFD Analysis Performed to Evaluate CSHL-Preliminary CFD Analysis Calculated CSHL of > 5 feet
- CSHLisaboveValueAllowedbyDesignSpecificationusedtoProcure
- CSHL is above Value Allowed by Design Specification used to Procure Strainer Assembly
- Hi gh CSHL Results from the existin g 5-inch and 5.5-inch Orifices used for ggFlow Balancing through 23 Strainer Modules 21 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
CSHL Reduction
- OptimalOrificeSizeDeterminedUsingCFDMethodbyIncrementally Optimal Orifice Size Determined , Using CFD Method , by Incrementally Adjusting Orifice Diameter over Multiple Model Executions
- Optimal Orifice Sizes Reduce Head Loss while Maintaining Flow Balance
- New Orifice Diameters Range from 6.5 inches to 8 inches
- NewCleanStrainerHeadLossis
<2feet New Clean Strainer Head Loss is 2 feet*Since Orifices are Holes Cut in Plenum Top Cover Plates, Replacement of Orifices Requires Replacement of Cover Plates
- Design Change for Plenum Cover Plate Replacement will be Issued in Support of Implementation and Completion during Fall 2012 Outage 22 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
CSHL Reduction
- PlenumCoverPlatesareshownbelow Plenum Cover Plates are shown below 23 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
Structural Qualification
- WBN,Unit1,StructuralCalculationCurrentlyQualifiestheStrainer WBN, Unit 1, Structural Calculation Currently Qualifies the Strainer Assembly for a Debris-Laden Head Loss of 3.65 feet
- Additional Preliminary Study Calculation Confirms WBN, Unit 1, Strainer is Structurally Qualified Up to a Head Loss of 5.7 feet
- New Total Head Loss is 3.83 feet, including Debris Loading at 120°F
-After Modifications Implemented 24 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal
- Min-KInsulationwasPrimarySourceofHeadLossduringRecent Min K Insulation was Primary Source of Head Loss during Recent Strainer Testing
-Test of Record, without Min-K, had Significantly Less Head Loss
- WBN will Remove All Min-K from Applicable Locations inside
Containment
-Min-K Insulation Currently used in Areas where Close Commodity Clearances Exist with Hot Piping 25 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal
-Kwillberemoved 15 Locations where Min K will be removed*Locations Distributed throughout Lower ContainmentWBNMitiStitCtlfIltiiCtitthhf
- WBN M a i n t a i ns St r i c t C on t ro l o f I nsu l a ti on i n C on t a i nmen t th roug h use o f Design Drawings
-Therefore, All Min-K Locations are known
- Walkdowns Performed during recent Spring 2011 Outage to Confirm Information on Drawings
- Volume of Min-K Varies from 0.03ft 3 to 0.94ft 3depending on Location 26 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal 27 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal 28 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal 29 WBN, Unit 1, Sump Strainer Structural Integrity (continued) Min-K Insulation Removal
- SomeLocationsmaybereplacedwithanIndustryApproved Some Locations may be replaced with an Industry Approved Non-Fibrous Type Insulation
-That does not Invalidate Sump Testing
- For Other Locations, if the Approved Material does not Provide Adequate Insulating Value, Modifications will be Required to Resolve Commodity ClearanceIssues Clearance Issues*Design Changes for Insulation Changes and Reroutes to Resolve Commodity Clearances will be issued in Support of Implementation and Completion during Fall 2012 Outage 30 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
SummaryExistingConfigurationModifiedConfiguration Existing Configuration Modified ConfigurationOrifice Diameters (inches)5.0 and 5.56.5, 7.0, 7.5, and 8.0Debris Head Loss (feet)*0.031.88CleanStrainerHeadLoss(feet)362195 Clean Strainer Head Loss (feet)3.62 1.95Total StrainerHead Loss (feet)3.653.83 Structural Qualification (feet)3.655.7**RHRNPSHMargin(feet) 94110*For120°FconditionsTemperatureCorrected2010TestHeadLossis109feetat190
°FMaximum RHR NPSH Margin (feet)9.4 11.0CS NPSH Margin (feet)5.534.74 For 120 F conditions. Temperature Corrected 2010 Test Head Loss is 1.09 feet at 190 F MaximumSump Temperature.** Value will Change Slightly as it is based on 5.0-and 5.5-inch Or ifice Diameters. New Value will Support Modified Total Strainer Head Loss.
31 WBN, Unit 1, Sump Strainer Structural Integrity (continued)
Summary*InitialDesignBasisDebrisLoadedThinBedTestResultedin Initial Design Basis Debris Loaded Thin Bed Test Resulted in Unacceptably High Head Loss
- High Test Head Loss will be resolved by 1.Implementing Design Change to Replace Plenum Cover Plate with Larger Orifice Sizes in Fall 2012 Outage 2.Completing Reanalysis of Structural Qualification of Strainer Assembly
with New Orifices 3.Im p lementin g Desi g n Chan ge to Remove Min-K Insulation in Fall 2012 pggg Outage 32 Applicability of Sump Strainer Structural Integrity Calculation to SQN/Plans for Insulation Remediation
- SQN does not Contain Fibrous Insulation
-No Insulation Remediation Required
- SQNIndividualStrainerModulesareSimilartoWBNUnit1StrainerModules
- SQN Individual Strainer Modules are Similar to WBN , Unit 1 , Strainer Modules-Similar Structural Margin Results Expected for Basic Module Elements
- SQN Strainer Stacks are Taller than WBN, Unit 1, Stacks
-72 inches versus 55 inches*SQN Plenum Assemblies are More Compact with Different Plenum Support Arrangement than WBN Plenum Assemblies
-123 inches X 69 inches -SQN versus 304 inches X 130 inches -WBN, Unit 1
- SQN Stack/Plenum Differences Prevent Direct Extrapolation of WBN, Unit 1, Results *SQN Configuration Specific Evaluations Required
- SQN Maximum Differential Pressure Remains below Current Analyzed Limit of 35feet 3.5 feet-Therefore, no plans for additional SQN structural margin assessments 33