U.S. Nuclear Regulatory Commission, Division of Systems Analysis
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The organization U.S. Nuclear Regulatory Commission, Division of Systems Analysis represents an institution, an association, or corporate body that is associated with resources found in University of Missouri-St. Louis Libraries.
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U.S. Nuclear Regulatory Commission, Division of Systems Analysis
Resource Information
The organization U.S. Nuclear Regulatory Commission, Division of Systems Analysis represents an institution, an association, or corporate body that is associated with resources found in University of Missouri-St. Louis Libraries.
- Label
- U.S. Nuclear Regulatory Commission, Division of Systems Analysis
- Subordinate unit
- Division of Systems Analysis
64 Items by the Organization U.S. Nuclear Regulatory Commission, Division of Systems Analysis
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Context of U.S. Nuclear Regulatory Commission, Division of Systems AnalysisContributor of
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- Rod bundle heat transfer facility : steady-state steam cooling experiments
- A quantitative impact assessment of hypothetical spent fuel reconfiguration in spent fuel storage casks and transportation packages
- An approach for validating actinide and fission product burnup credit criticality safety analyses--criticality (k[subscript eff]) predictions
- An approach for validating actinide and fission product burnup credit criticality safety analyses--isotopic composition predictions
- Analysis with TRACE code of ROSA test 1.1 : ECCS water injection under natural circulation condition
- Assessment against ACHILLES reflood experiment with TRACE V5.0 Patch3
- Assessment of TRACE 5.0 against ROSA-2 test 3 countertest to PKL
- Assessment of TRACE V5.0.Patch 4 code against PWR PACTEL loop seal clearing experiment
- Assessment of critical subcooled flow through cracks in large and small pipes using TRACE and RELAP5
- Assessment of the wall film condensation model with non-condensable gas in RELAP5 and TRACE for vertical tube and plate geometries
- Axial moderator density distributions, control blade usage, and axial burnup distributions for extended BWR burnup credit
- BEPU analysis and benchmark with IIST 2% SBLOCA experiment using TRACE/DAKOTA
- BWR ECCS pump suction concerns following a LOCA
- BWR anticipated transients without scram in the MELLLA+ expanded operating domain
- Benchmarking of a generic CANDU reactor with PARCS, MCNP and RFSP
- Cladding behavior during postulated loss-of-coolant accidents
- Comparison of the U.S. NRC PARCS core neutronics simulator against in-core detector measurements for LWR applications
- Computational benchmark for estimation of reactivity margin from fission products and minor actinides in BWR burnup credit
- Core exit temperature response during an SBLOCA event in the Ascó NPP
- Development of a coupled TRACE/PARCS model for KKL and benchmark against the turbine trip test
- Evaluation of TRACE spacer grid model with FLECHT-SEASET reflood test
- Feedwater line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- IBLOCA analysis for Vandellós-NPP using RELAP5/MOD3.3 sensitivity calculations to EOP actions
- Implementation of advanced multigroup nodal and pin power reconstruction methods into PARCS 3.1
- Investigation of the loop seal clearing phenomena for the ATLAS DVI line and cold leg SBLOCA tests using MARS-KS and RELAP5/MOD3.3
- Laminar hydraulic analysis of a commercial pressurized water reactor fuel assembly
- Loss of flow analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP
- Main stream line break analysis for lungmen ABWR
- Post-test analysis of ROSA-2 test 2 (IBLOCA) with TRACE
- Post-test analysis of cold leg small break 4.1% at PSB-VVER facility using TRACE V5.0
- Post-test calculation of the PKL-2 test G7.1 using RELAP5/MOD3.3
- Post-test calculation of the ROSA-2 test 3 using RELAP5/MOD3.3
- Post-test thermal-hydraulic analysis of PKL tests F1.1 and F1.2
- Proposal for the development and implementation of an uncertainty and sensitivity analysis module in SNAP
- RELAP5 analysis of mitigation strategy for extended blackout power condition in PWR
- RELAP5 and TRACE calculations of LOCA in PWR
- RELAP5 model of a CANDU-6 (Embalse) nuclear power plant : application to a turbine trip event
- RELAP5/MOD3.3 analysis of event with actuation of safety injection system at the loss of external power
- RELAP5/MOD3.3 model assessment of Maanshan Nuclear Power Plant with SNAP interface
- Research reactor 'MARIA' primary cooling loop transient analysis using RELAP5 Mod 3.3
- Rod bundle heat transfer facility steam cooling with droplet injection experiments data report
- Rod bundle heat transfer facility two-phase mixture level swell and uncovery test experiments data report
- SNAP/RADTRAD 4.0 : description of models and methods
- Semiscale S-NC-02 and S-NC-03 natural circulation tests performed by RELAP5/MOD3.3 Patch05
- Sensitivity analyses of a hypothetical 6 inch break, LOCA in Ascó NPP using RELAP/MOD3.2
- Simulation of LSTF hot leg break (OECD/NEA ROSA-2 test 1) with TRACE code : application to a PWR NPP model
- Simulation of LSTF upper head break (OECD/NEA ROSA test 6.1) with TRACE code : application to a PWR NPP model
- Simulation of ROSA-2 test-2 experiment : application to nuclear power plant
- Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL facility using TRACE 5
- Simulation of the PKL-G7.1 experiment in a Westinghouse nuclear power plant using RELAP5/MOD3.3
- Steam line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- TRACE VVER-1000/V-320 model validation
- TRACE VVER-440/V-213 model validation
- TRACE analysis on heat removal decrease accidents for AP1000
- TRACE assessment for effect of spacer grid in RBHT reflood heat transfer experiments
- TRACE/RELAP5 comparative calculations for double-ended LBLOCA and SBO
- TRACE/SNAP model establishment of Chinshan nuclear power plant for ultimate response guideline
- Technical basis for the containment protection and release reduction rulemaking for boiling water reactors with Mark I and Mark II containments
- The alternate mitigation strategies study of Chinshan BWR/4 by using the LOCA and SBO analysis of TRACE
- The analysis and study of ELAP event and mitigation strategies using TRACE code for Maanshan PWR
- Thermal hydraulic and fuel rod mechanical combination analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD3.3/FRAPTRAN/Python in SNAP interface
- Using SNAP/RADTRAD and HABIT to establish the analysis methodology for Maanshan PWR
- Using TRACE, MELCOR, CFD, and FRAPTRAN to establish the analysis methodology for Chinshan Nuclear Power Plant spent fuel pool
- Validation of RELAP5 model of ringhals 4 against a load step test at uprated power
Issuing body of
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No enriched resources found
- Semiscale S-NC-02 and S-NC-03 natural circulation tests performed by RELAP5/MOD3.3 Patch05
- Laminar hydraulic analysis of a commercial pressurized water reactor fuel assembly
- Validation of RELAP5 model of ringhals 4 against a load step test at uprated power
- Loss of flow analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP
- Main stream line break analysis for lungmen ABWR
- TRACE assessment for effect of spacer grid in RBHT reflood heat transfer experiments
- Assessment of the wall film condensation model with non-condensable gas in RELAP5 and TRACE for vertical tube and plate geometries
- Post-test analysis of ROSA-2 test 2 (IBLOCA) with TRACE
- Assessment of critical subcooled flow through cracks in large and small pipes using TRACE and RELAP5
- Post-test analysis of cold leg small break 4.1% at PSB-VVER facility using TRACE V5.0
- Post-test calculation of the PKL-2 test G7.1 using RELAP5/MOD3.3
- The analysis and study of ELAP event and mitigation strategies using TRACE code for Maanshan PWR
- Post-test calculation of the ROSA-2 test 3 using RELAP5/MOD3.3
- Simulation of ROSA-2 test-2 experiment : application to nuclear power plant
- Investigation of the loop seal clearing phenomena for the ATLAS DVI line and cold leg SBLOCA tests using MARS-KS and RELAP5/MOD3.3
- Using SNAP/RADTRAD and HABIT to establish the analysis methodology for Maanshan PWR
- Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL facility using TRACE 5
- Implementation of advanced multigroup nodal and pin power reconstruction methods into PARCS 3.1
- RELAP5 analysis of mitigation strategy for extended blackout power condition in PWR
- TRACE/RELAP5 comparative calculations for double-ended LBLOCA and SBO
- RELAP5 and TRACE calculations of LOCA in PWR
- TRACE/SNAP model establishment of Chinshan nuclear power plant for ultimate response guideline
- Simulation of the PKL-G7.1 experiment in a Westinghouse nuclear power plant using RELAP5/MOD3.3
- RELAP5 model of a CANDU-6 (Embalse) nuclear power plant : application to a turbine trip event
- Steam line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- RELAP5/MOD3.3 analysis of event with actuation of safety injection system at the loss of external power
- Thermal hydraulic and fuel rod mechanical combination analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD3.3/FRAPTRAN/Python in SNAP interface
- RELAP5/MOD3.3 model assessment of Maanshan Nuclear Power Plant with SNAP interface
- IBLOCA analysis for Vandellós-NPP using RELAP5/MOD3.3 sensitivity calculations to EOP actions
- Using TRACE, MELCOR, CFD, and FRAPTRAN to establish the analysis methodology for Chinshan Nuclear Power Plant spent fuel pool
- Feedwater line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- Evaluation of TRACE spacer grid model with FLECHT-SEASET reflood test
- Development of a coupled TRACE/PARCS model for KKL and benchmark against the turbine trip test
- TRACE VVER-1000/V-320 model validation
- Core exit temperature response during an SBLOCA event in the Ascó NPP
- TRACE VVER-440/V-213 model validation
- Assessment of TRACE 5.0 against ROSA-2 test 3 countertest to PKL
- BWR ECCS pump suction concerns following a LOCA
- Assessment of TRACE V5.0.Patch 4 code against PWR PACTEL loop seal clearing experiment
Publisher of
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No enriched resources found
Sponsoring body of
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No enriched resources found
- Post-test thermal-hydraulic analysis of PKL tests F1.1 and F1.2
- Simulation of LSTF upper head break (OECD/NEA ROSA test 6.1) with TRACE code : application to a PWR NPP model
- Benchmarking of a generic CANDU reactor with PARCS, MCNP and RFSP
- BWR anticipated transients without scram in the MELLLA+ expanded operating domain
- Cladding behavior during postulated loss-of-coolant accidents
- Comparison of the U.S. NRC PARCS core neutronics simulator against in-core detector measurements for LWR applications
- SNAP/RADTRAD 4.0 : description of models and methods
- Simulation of LSTF hot leg break (OECD/NEA ROSA-2 test 1) with TRACE code : application to a PWR NPP model
- Computational benchmark for estimation of reactivity margin from fission products and minor actinides in BWR burnup credit
- Rod bundle heat transfer facility two-phase mixture level swell and uncovery test experiments data report
- Assessment against ACHILLES reflood experiment with TRACE V5.0 Patch3
- The alternate mitigation strategies study of Chinshan BWR/4 by using the LOCA and SBO analysis of TRACE
- Rod bundle heat transfer facility steam cooling with droplet injection experiments data report
- BEPU analysis and benchmark with IIST 2% SBLOCA experiment using TRACE/DAKOTA
- Axial moderator density distributions, control blade usage, and axial burnup distributions for extended BWR burnup credit
- A quantitative impact assessment of hypothetical spent fuel reconfiguration in spent fuel storage casks and transportation packages
- Rod bundle heat transfer facility : steady-state steam cooling experiments
- TRACE analysis on heat removal decrease accidents for AP1000
- Research reactor 'MARIA' primary cooling loop transient analysis using RELAP5 Mod 3.3
- An approach for validating actinide and fission product burnup credit criticality safety analyses--isotopic composition predictions
- Proposal for the development and implementation of an uncertainty and sensitivity analysis module in SNAP
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<div class="citation" vocab="http://schema.org/"><i class="fa fa-external-link-square fa-fw"></i> Data from <span resource="http://link.umsl.edu/resource/E5jfVu2xgt0/" typeof="Organization http://bibfra.me/vocab/lite/Organization"><span property="name http://bibfra.me/vocab/lite/label"><a href="http://link.umsl.edu/resource/E5jfVu2xgt0/">U.S. Nuclear Regulatory Commission, Division of Systems Analysis</a></span> - <span property="potentialAction" typeOf="OrganizeAction"><span property="agent" typeof="LibrarySystem http://library.link/vocab/LibrarySystem" resource="http://link.umsl.edu/"><span property="name http://bibfra.me/vocab/lite/label"><a property="url" href="http://link.umsl.edu/">University of Missouri-St. Louis Libraries</a></span></span></span></span></div>
