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Category:Meeting Briefing Package/Handouts
MONTHYEARML20184A2872020-07-0101 July 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and General Design Criterion 33, PM-0720-70785, Revision 0 ML20150C5172020-05-29029 May 2020 LLC Submittal of Presentation Materials Entitled NRC Public Meeting Presentation: Boron Redistribution and Associated Design and DCA Changes, PM-0620-70336, Revision 0 ML20150C8812020-05-29029 May 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and Associated Design and DCA Changes, PM-0620-70244, Revision 0 ML20099H0802020-04-0808 April 2020 LLC - Submittal of Presentation Materials Entitled NRC Public Meeting: Revisions to Nuscale'S EPZ Sizing Methodology Topical Report, PM-0420-69598, Revision 0 ML20094H6742020-04-0303 April 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation NuScale Topic-Probabilistic Risk Assessment with a Focus on Emergency Core Cooling System Analysis PM-0420-69559, Revision 0 ML20072H3272020-03-0909 March 2020 LLC - Public Meeting Presentation: Topic - Emergency Core Cooling System Boron (ECCS) Distribution, PM-0320-69218, Revision 0 ML20069A9692020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, Revision O ML20069A1632020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0 ML20069A1542020-03-0404 March 2020 ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0 ML20066G2802020-03-0303 March 2020 LLC, Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Topic - Hydrogen Monitoring, PM-0220-69071, Revision 0 ML20066G2772020-02-28028 February 2020 ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, Revision 0 ML20054A9622020-02-19019 February 2020 ACRS Subcommittee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0 ML20054B6342020-02-17017 February 2020 Nuscale - ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0 ML20054A4062020-02-17017 February 2020 ACRS Subcommittee Presentation: NuScale Topical Report- Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0 ML20035C7892020-02-0404 February 2020 Nuscale - ACRS Subcommittee Presentation, FSAR, Steam Generator Design ML19340B5912019-12-0606 December 2019 Public Meeting - NuScale Incorporated by Reference (Ibr) Slides ML19324C8362019-11-13013 November 2019 LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Source Term Methodology Application, PM-1119-67928, Revision 0 ML19204A2772019-07-22022 July 2019 Discussion Topic July 22, 2019 Public Meeting to Discuss Interface Requirement Associated with the NuScale Design Certification Application ML19190A0882019-07-0303 July 2019 LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation, NuScale FSAR Chapter 20, Mitigation of Beyond-Design-Basis Events, PM-0719-66189, Revision 0 ML19169A1662019-06-14014 June 2019 Mitigation of Beyond-Design-Basis-Events, NuScale FSAR Chapter 20, PM-0619-65964, Revision 0 ML19135A0642019-05-0909 May 2019 LLC - Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0 ML19134A0742019-05-0909 May 2019 LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0 ML19105A1352019-04-11011 April 2019 ACRS Presentation: NuScale Chapter 5, Reactor Coolant System and Connecting Systems Overview, PM-0419-65159, Revision 0 ML19098A7802019-04-0202 April 2019 LLC - Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Chapter 9, Auxiliary Systems, PM-0319-64807, Revision 0 ML19079A2382019-03-20020 March 2019 Enclosure 1: ACRS Presentation: NuScale Chapter 9, Auxiliary Systems, PM-0319-64807, Revision 0 ML19073A0482019-03-11011 March 2019 LLC - ACRS Presentation Chapter 10 - Steam and Power Conversion System, PM-0219-64501, Revision 1 ML19073A0512019-03-11011 March 2019 ACRS Presentation Chapter 11 - Radioactive Waste Management, PM-0219-64543, Revision 1 ML19050A1632019-02-20020 February 2019 Enclosure 1: ACRS Presentation Chapter 11 - Radioactive Waste Management, PM-0219-64543, Revision 0 ML19053A5182019-02-19019 February 2019 Nuscale - Dhrs and ECCS Logic Changes, PM-0219-64563, Revision 0 ML19045A3062019-02-12012 February 2019 PM-0219-64475-NP, Revision 0, Reactor Vessel Flange Tool (Rft) Design Plan and Schedule. ML19042A0732019-02-0505 February 2019 Nuscale - Chapters 2 and 17 ACRS Full Committee, PM-0219-64333, Revision 0 ML19024A2042019-01-23023 January 2019 ECCS Valve DCA Demonstration Testing, PM-0119-64160-NP, Revision 0 ML18362A1422018-12-19019 December 2018 Enclosure 2: Absorption Coefficient for Pool Water and Reactor Building Floor Interaction, PM-1218-63882, Revision 0 ML18347A2032018-12-18018 December 2018 Enclosure 1: ACRS Presentation Chapter 17 - Quality Assurance and Reliability Assurance, PM-1218-63732, Revision 0 ML18347A2072018-12-18018 December 2018 ACRS Presentation Chapter 2 - Site Characteristics and Site Parameters, PM-1218-63753, Revision 0 ML18331A1602018-11-20020 November 2018 Enclosure 2: Addressing Component Fatigue for NuScale DCA, November 20, 2018 ML18338A1032018-11-14014 November 2018 Nuscale - LTC RAI Closure Plan, PM-1118-63594-NP, Revision 0, Nonproprietary Version ML18235A3152018-08-23023 August 2018 LLC - Presentation Materials Entitled ACRS Presentation: NuScale Instrumentation and Controls Design Overview, PM-0618-60212, Revision 0 ML18222A1932018-08-0808 August 2018 Enclosure 1: NuScale Accident Source Terms Methodology Options PM-0818-61166, Revision 0 ML18213A1572018-08-0101 August 2018 Risk-Informed Review of NuScale Initial Test Program, PM-0718-61123, Revision 0 ML18220A8742018-07-25025 July 2018 Enclosunuscale CIRLT-RAI 9474 ML18205A3182018-07-24024 July 2018 NRC Staff Questions for the August 1, 2018 Public Meeting with NuScale on ITP ML18116A4422018-04-24024 April 2018 Enclosure 2: NuScale Power Comprehensive Vibration Assessment Program - Audit Report Summary Discussion, Revision 0, PM-0218-58803-NP (Non-Proprietary Version) ML18090A0032018-04-0202 April 2018 Handout for the April 2, 2018 Public Meeting to Discuss NuScale Plans to Respond to the NRC Staff RAIs Related to Sections 3.7 and 3.8 ML18085A0722018-03-26026 March 2018 NRC Staff Questions for April 9, 2018 Public Meeting with NuScale on DCA Chapter 20 ML18037A6732018-02-0808 February 2018 PM-0218-58480, Revision 0, Shutdown Capability of the NuScale Power Module. ML18019A1612018-01-23023 January 2018 LLC - Shutdown Capability of the NuScale Power Module ML18019A1632018-01-23023 January 2018 Source Term Revision, Revision 0, PM-0118-58201 2020-07-01
[Table view] Category:Slides and Viewgraphs
MONTHYEARML20184A2872020-07-0101 July 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and General Design Criterion 33, PM-0720-70785, Revision 0 ML20150C8812020-05-29029 May 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Topic - Boron Redistribution and Associated Design and DCA Changes, PM-0620-70244, Revision 0 ML20150C5172020-05-29029 May 2020 LLC Submittal of Presentation Materials Entitled NRC Public Meeting Presentation: Boron Redistribution and Associated Design and DCA Changes, PM-0620-70336, Revision 0 ML20133J9142020-05-11011 May 2020 LLC Submittal of Presentation Materials Entitled Public Meeting Presentation: Topic - Emergency Core Cooling System (ECCS) Boron Distribution, PM-0420-69511, Revision 0 ML20099H0802020-04-0808 April 2020 LLC - Submittal of Presentation Materials Entitled NRC Public Meeting: Revisions to Nuscale'S EPZ Sizing Methodology Topical Report, PM-0420-69598, Revision 0 ML20097F1922020-04-0606 April 2020 Nuscale Power, LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: Nuscale Topic - Hydrogen/Oxygen Monitoring, PM-0420-69518, Revision 0 ML20094H6742020-04-0303 April 2020 LLC Submittal of Presentation Materials Entitled ACRS Full Committee Presentation NuScale Topic-Probabilistic Risk Assessment with a Focus on Emergency Core Cooling System Analysis PM-0420-69559, Revision 0 ML20072H3272020-03-0909 March 2020 LLC - Public Meeting Presentation: Topic - Emergency Core Cooling System Boron (ECCS) Distribution, PM-0320-69218, Revision 0 ML20069A1632020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0 ML20069A9692020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, Revision O ML20069A1542020-03-0404 March 2020 ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0 ML20066G2802020-03-0303 March 2020 LLC, Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Topic - Hydrogen Monitoring, PM-0220-69071, Revision 0 ML20066F4672020-02-28028 February 2020 Enclosure 1: ACRS Subcommittee Presentation: NuScale FSAR Topic- Resolution of Chapter 15 Phase 2 Open Items, PM-0220-69062, Revision 0 ML20066G2772020-02-28028 February 2020 ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, Revision 0 ML20054A9622020-02-19019 February 2020 ACRS Subcommittee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0 ML20054A4062020-02-17017 February 2020 ACRS Subcommittee Presentation: NuScale Topical Report- Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0 ML20054B6342020-02-17017 February 2020 Nuscale - ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0 ML20035C7892020-02-0404 February 2020 Nuscale - ACRS Subcommittee Presentation, FSAR, Steam Generator Design ML19340B5912019-12-0606 December 2019 Public Meeting - NuScale Incorporated by Reference (Ibr) Slides ML19324C8362019-11-13013 November 2019 LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Source Term Methodology Application, PM-1119-67928, Revision 0 ML19192A1662019-07-10010 July 2019 Presentation Entitled, Evaluation Methodology for Stability Analysis of the NuScale Power Module. ML19169A1662019-06-14014 June 2019 Mitigation of Beyond-Design-Basis-Events, NuScale FSAR Chapter 20, PM-0619-65964, Revision 0 ML19135A0642019-05-0909 May 2019 LLC - Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0 ML19105A1352019-04-11011 April 2019 ACRS Presentation: NuScale Chapter 5, Reactor Coolant System and Connecting Systems Overview, PM-0419-65159, Revision 0 ML19105A1312019-04-10010 April 2019 Enclosuacrs Presentation: Chapter 4, Reactor Overview, PM-0419-65096, Revision 0 ML19098A7802019-04-0202 April 2019 LLC - Submittal of Presentation Materials Entitled ACRS Full Committee Presentation: NuScale Chapter 9, Auxiliary Systems, PM-0319-64807, Revision 0 ML19073A0482019-03-11011 March 2019 LLC - ACRS Presentation Chapter 10 - Steam and Power Conversion System, PM-0219-64501, Revision 1 ML19073A0512019-03-11011 March 2019 ACRS Presentation Chapter 11 - Radioactive Waste Management, PM-0219-64543, Revision 1 ML19073A0532019-03-11011 March 2019 Enclosure 1: ACRS Presentation Chapter 12 - Radiation Protection, PM-0219-64534, Revision 1 ML19050A1632019-02-20020 February 2019 Enclosure 1: ACRS Presentation Chapter 11 - Radioactive Waste Management, PM-0219-64543, Revision 0 ML19053A5182019-02-19019 February 2019 Nuscale - Dhrs and ECCS Logic Changes, PM-0219-64563, Revision 0 ML19050A3982019-02-13013 February 2019 Enclosure 1: ACRS Presentation Chapter 10 - Steam and Power Conversion System, PM-0219-64501, Revision 0 ML19045A3062019-02-12012 February 2019 PM-0219-64475-NP, Revision 0, Reactor Vessel Flange Tool (Rft) Design Plan and Schedule. ML19042A0732019-02-0505 February 2019 Nuscale - Chapters 2 and 17 ACRS Full Committee, PM-0219-64333, Revision 0 ML19024A2042019-01-23023 January 2019 ECCS Valve DCA Demonstration Testing, PM-0119-64160-NP, Revision 0 ML18360A1782018-12-19019 December 2018 Enclosure 2: Presentation Titled, Addressing Component Fatigue for NuScale Dca. ML18362A1422018-12-19019 December 2018 Enclosure 2: Absorption Coefficient for Pool Water and Reactor Building Floor Interaction, PM-1218-63882, Revision 0 ML18347A2032018-12-18018 December 2018 Enclosure 1: ACRS Presentation Chapter 17 - Quality Assurance and Reliability Assurance, PM-1218-63732, Revision 0 ML18347A2072018-12-18018 December 2018 ACRS Presentation Chapter 2 - Site Characteristics and Site Parameters, PM-1218-63753, Revision 0 ML19025A2212018-12-18018 December 2018 Enclosure 2: Slides Titled, Basis for an Rft Stand COL Item, PM-1218-63892-NP, Revision 0, Non-Proprietary Version ML18331A1602018-11-20020 November 2018 Enclosure 2: Addressing Component Fatigue for NuScale DCA, November 20, 2018 ML18338A1032018-11-14014 November 2018 Nuscale - LTC RAI Closure Plan, PM-1118-63594-NP, Revision 0, Nonproprietary Version ML18235A3152018-08-23023 August 2018 LLC - Presentation Materials Entitled ACRS Presentation: NuScale Instrumentation and Controls Design Overview, PM-0618-60212, Revision 0 ML18222A1932018-08-0808 August 2018 Enclosure 1: NuScale Accident Source Terms Methodology Options PM-0818-61166, Revision 0 ML18213A1572018-08-0101 August 2018 Risk-Informed Review of NuScale Initial Test Program, PM-0718-61123, Revision 0 ML18220A8742018-07-25025 July 2018 Enclosunuscale CIRLT-RAI 9474 ML18156A1262018-06-0101 June 2018 LLC Submittal of Presentation Materials Entitled, ACRS Presentation Chapter 8 Overview, for Use During a Public Meeting on June 6, 2018 ML18156A1272018-05-31031 May 2018 LLC Submittal of Presentation Materials Entitled NuScale Power Containment Leak Rate Test Program, PM-0518-59978, Revision 1 ML18116A4422018-04-24024 April 2018 Enclosure 2: NuScale Power Comprehensive Vibration Assessment Program - Audit Report Summary Discussion, Revision 0, PM-0218-58803-NP (Non-Proprietary Version) ML18037A6732018-02-0808 February 2018 PM-0218-58480, Revision 0, Shutdown Capability of the NuScale Power Module. 2020-07-01
[Table view] Category:Topical Report
MONTHYEARML20099H0802020-04-0808 April 2020 LLC - Submittal of Presentation Materials Entitled NRC Public Meeting: Revisions to Nuscale'S EPZ Sizing Methodology Topical Report, PM-0420-69598, Revision 0 ML20069A1632020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0 ML20069A9692020-03-0505 March 2020 ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, Revision O ML20069A1542020-03-0404 March 2020 ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0 ML19352G6972019-12-18018 December 2019 LLC Submittal of Long-Term Cooling Methodology, TR-0916-51299, Revision 2 ML19319C7872019-11-15015 November 2019 LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 3 ML19324C8392019-11-13013 November 2019 LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation NuScale Topical Report -Accident Source Term Methodology, PM-1119-67927, Revision 0 ML19211C9012019-07-30030 July 2019 LLC Submittal of NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-0816-51127, Revision 2 ML19192A1662019-07-10010 July 2019 Presentation Entitled, Evaluation Methodology for Stability Analysis of the NuScale Power Module. ML19175A2992019-06-24024 June 2019 LLC Submittal of NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, TR-1010-859-NP, Revision 4 ML19136A3322019-05-16016 May 2019 LLC Submittal of Changes to the Pipe Rupture Hazards Analysis Technical Report, TR-0818-61384, Revision 1 ML19087A2882019-03-28028 March 2019 LLC Updates to Final Safety Analysis Report, Section 1.6,Materials Referenced, Section 3.6, Protection Against Dynamic Effects Associated with Postulated Rupture of Piping, and the Pipe Rupture Hazards Analysis, TR-0818-61384 ML18354B4002018-12-20020 December 2018 LLC - Submittal of Pipe Rupture Hazards Analysis, TR-0818-61384, Revision 1 ML18360A6322018-12-19019 December 2018 LLC - Submittal of Approved Version of Topical Report TR-0116-21012, NuScale Power Critical Heat Flux Corrections, Revision 1 ML18341A3372018-12-0707 December 2018 LLC Submittal of NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, TR-0918-60894, Revision 0 ML18338A0312018-12-0404 December 2018 Final Safety Evaluation for NuScale Power, TR-0915-17564-P, Subchannel Analyses Methodology, Revision 2 ML18310A1542018-11-0505 November 2018 LLC Submittal of Fuel Storage Rack Analysis, TR-0816-49833, Revision 1 ML18096B8852018-04-0606 April 2018 LLC Submittal of Technical Report TR-1015-18177, Pressure and Temperature Limits Methodology, Revision 1 2020-04-08
[Table view] |
Text
L0-0320-69151
Enclosure:
"ACRS Full Committee Presentation: NuScale Topical Report- Rod Ejection Accident Methodology," PM-0320-69146, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., SJite 200 Corvallis, 0-egon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Nonproprietary ACRS Full Committee Presentation N uScale Topical Report Rod Ejection Accident Methodology March 5, 2020 PM-0320-69146
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Template# : 0000.21727-F0 1 R5
Presenters Kenny Anderson Nuclear Fuels Analyst Matthew Presson Licensing Project Manager 2
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Opening Remarks - NuScale T/H Methods NRELAP5 System T/H Analysis Basis code TR-0516-49422-P I
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I
- NRELAP5 code developed from RELAP5-3D 1---;.i: :
LOCA EM
- Valve opening event I I I I TR-0516-49084-P Modified to address NuScale- *-------------------------* ,--------------
I' II
- Containment :
specific phenomena/systems :----- ;-------: 1 - - - - - - - -~, respo nse :
I I ~ analysis :
,I Control rod *I TR-0516-49416-P
- LOCA Evaluation Model (EM) : ejection : I I-----------*** I I
I developed following RG 1.203 EMDAP : (T/H response)
I 0 1--_ :
I Non-LOCA I
L--------------' i' EM
LOCA EM extended to derive TR-0716-50350-P EMs for other events as shown in this figure. FSARCh 5, RAI 9508 LOCA EM assessment basis --------------
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- Extended :
leveraged for non-LOCA. ~ DHRS :
- cooli ng :
- Additional supporting EMs include
FSARCh 15
- Nuclear Analysis Codes - ----*-----*---
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I TR-0716-50350-P-A ' - - - - -~*,* Overcooling
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, power :
Critical Heat Flux - I I
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Subchannel Analysis - : j Long tefm :
TR-0915-17564-P-A ' - - - - - - - - - - - . .;,iI cooling with :
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Agenda
- Steady State Initialization
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Template#: 0000.21727-F01 R5
Overview
- NuScale seeks approval of methodology for modeling rod ejection accident (REA) events
- REA is unique in comparison to other Ch. 15 events Description Rod Ejection Other Events Dominant Physics Nuclear Thermal-Hydraulics Timing milli-sec secto hr Spatially Local Global Peak power -5x Full Power -1 .2x Full Power Integrated Energy Low Lowto High Failure of ASME Class 1 Postulated Cause Single Equipment Failure Pressure Boundary Acceptance Criteria Specialized Generic 5
PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Unique Event Acceptance Criteria Topical Criteria Description Unique?
Section Maximum reactor coolant system pressure 5.3 No Hot zero power (HZP) fuel cladding failure 5.5.2 Yes FGR effect on cladding differential pressure N/A Yes Critical heat flux (CHF) fuel cladding failure 5.4.1 No Cladding oxidation-based PCMI failure 5.5.3 Yes Cladding excess hydrogen-based PCMI failure N/A Yes Incipient fuel melting cladding failure 5.5.1 No Peak radial average fuel enthalpy for core cooling 5.5.2 Yes Fuel melting for core cooling 5.5.1 No Fission product inventory (failed fuel census) 5.6 Yes
- Submitted NuScale design and method inherently precludes fuel failure, thus no accident radiological consequences are evaluated.
- PCMI: Pellet-Clad Mechanical Interaction 6
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Revised PCMI Criteria
- In general, the NuScale REA methodology has adopted the limiting criteria of the 'Clifford Letter'(ML14188C423), now included in draft guide DG-1327 (ML16124A200). In spirit, NuScale is prepared forthis regulatory change:
- Closed session presents example results , showing large margins for enthalpy rise
- A techn ical 'formality' inhibits complete adoption at this time. NuScale does not currently have a val idated cladding H2 model to convert local exposure to excess cladding hydrogen
- Oxidation criteria from NUREG-0800 Section 4.2 , Appendix B (ML07074000) is used
- To simplify method , no exposure is credited (Limit: 75 Li cal/gm)
- NuScale MS cladding less susceptible than other zirc alloy-type clad used in the industry 200 200 - - - - - - - - - - - - - - - - - ,
175 (0.04, 150) ~ (0, lSO) (75. 150)
~
150 }1~ - - ~ ~ ~ -
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"; 125 Cladding Failure t 125 Oadding Failure II)
~ 100 IJ 100
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-5C (0.08, 75) ;!:
w 75 ~
75 Cladding Intact ci
- << 50 LI. 50 i (0.20, 0) l.
25 25 0
0 0 so 100 150 200 250 0 0.04 0.08 0.12 0.16 0.2 txcau a..ddlnt Hydn>sen (wppm)
Oxide/Wall Thickness
._, NUSCALE .
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Tem pl ate#: 0000-21 727-F01 R5
Unique Event Method (Flowchart)
NRELAPS Dynamic System Response SIMULATES Steady State Initialization S1MULATE-3K Dynamic Core
Response
VIPRE-01 SubchannelCHF Evaluation MCHFR Above Correlation Design Limit 0
Adiabatic Heatup Fuel Response 8
PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Steady-State Initialization
- SIMULATES: Setup the core response analysis
- Code shown to be appropriate in TR-0616-48793-A (Nuclear Analysis Codes and Methods Qualification)
- Determination of the worst rod stuck out (WRSO)
- Assumption bounds potential for ejected assembly to damage adjacent control rod assembly
- Due to rapid nature of the event, location does not significantly affect the results in NuScale application 9
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Dynamic Core Response
- Benchmarked to SPERT-111 experiment and NEACRP computational benchmark
- Benchmarks demonstrate the combined transient neutronic, thermal-hydraulic, and fuel pin modeling capabilities
- S1MULATE-3K results generally in excellent agreement with the results from the two benchmark problems
- Uncertainties applied for each simulation:
- Delayed Neutron Fraction
- Ejected Rod Worth
- Doppler Temperature Coefficient
- Moderator Temperature Coefficient 10 PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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CHF Evaluation
- VIPRE-01: Model detailed thermal-hydraulics
- Evaluate critical heat flux (CHF) acceptance c ~---------____,
- Code shown to be appropriate in TR-0915-17.
Analysis Methodology)
- Unique event differences in method:
- Smaller axial nodalization (smaller time steps)
- Radial power distribution (case-specific)
- Axial power distribution (peak assembly)
- Convergence parameters
- Additional parametric sensitivity cases pe application to holistically justify difference 11 PM-0320-691 46
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Templ ate#: 0000-21727-F01 R5
Adiabatic Fuel Heatup
- Hand-Calculation: Model fuel response
- Total energy (from S1MULATE-3K) during the transient is integrated
- Conservative as no energy is allowed to leave the fuel rod
- Energy is then converted into either a temperature or enthalpy increase
- Fuel rod geometry, heat capacity, and power peaking factors taken into account
- Calculated values compared to NRC developed acceptance criteria
- Example values provided in closed session 12 PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Dynamic System Response I
- NRELAPS: Evaluate system response for input to CHF Evaluation CON1IOI. ROD ,
DRIVE MECHANISM
- Code shown to be appropriate in TR-0516-49416 (Non-LOCA Methodologies)
- Transient power from S1MULATE-3K MAit STEAM RISER utilized as input (PRIMARY R.OW)
STEAIJ GENERATOR
- No reactivity calculation performed in NRELAPS (SECON DARY FlOWI
- Provides system thermal-hydraulic CONTAINMENT 1/ESSB.
conditions to subchannel (CHF) evaluation
- System flow, pressure, and inlet temperature REAClOR PRES$URE vessa.
CORE
- 'Screens' cases for potential to be limiting (PRIIARY FLOW)
- Family of limiting cases evaluated with VIPRE-01 13 PM-0320-69146
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Template#: 0000.21727-F01 RS
Dynamic System Response II
- NRELAPS: Evaluate system response for pressurization
- Limiting scenario: Low ejected worth that raises the power quickly to just below both the high power and high power rate trip 'setpoints'
- Point-kinetics model used based on bounding static worth
- Peak system pressure calculated compared to acceptance criteria
- Example results to be presented in closed session 14 PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Summary
- A conservative analysis method for the unique rod ejection accident
- Topical report provides details and justification for:
- Software tools and acceptance criteria used
- Applicability of the method and tools
- Appropriate treatment of uncertainties
- Results from application of the method provide input to FSAR Chapter 15 15 PM-0320-69146
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Acronyms
- GDC-General Design Criteria
- MCHFR- Minimum Critical Heat Flux Ratio
- NEACRP- Nuclear Energy Agency Committee on Reactor Physics
- PCMI - Pellet Clad Mechanical Interaction
- REA - Rod Ejection Accident
- RIA- Reactivity Initiated Accident
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