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 Biddle Law Library- University of Pennsylvania Law School.
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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 Biddle Law Library- University of Pennsylvania Law School.
- Label
- U.S. Nuclear Regulatory Commission, Division of Systems Analysis
- Subordinate unit
- Division of Systems Analysis
53 Items by the Organization U.S. Nuclear Regulatory Commission, Division of Systems Analysis
Context
Context of U.S. Nuclear Regulatory Commission, Division of Systems AnalysisContributor of
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- Evaluation of TRACE spacer grid model with FLECHT-SEASET reflood test
- Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code
- Analysis with TRACE code of ROSA test 1.1 : ECCS water injection under natural circulation condition
- Analysis with TRACE code of Rosa test 1.2 : small LOCA in the hot-leg with HPI and accumulator actuation
- 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
- BEPU analysis and benchmark with IIST 2% SBLOCA experiment using TRACE/DAKOTA
- Benchmarking of a generic CANDU reactor with PARCS, MCNP and RFSP
- Comparison of the U.S. NRC PARCS core neutronics simulator against in-core detector measurements for LWR applications
- 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
- Feedwater line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- Fuel rod behavior and uncertainty analysis by FRAPTRAN/TRACE/DAKOTA code in maanshan LBLOCA
- 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
- 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
- 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/RELAP5 comparative calculations for double-ended LBLOCA and SBO
- TRACE/SNAP model establishment of Chinshan nuclear power plant for ultimate response guideline
- 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
- Uncertainty and sensitivity investigations with TRACE-SUSA and TRACE-DAKOTA by means of post-test calculations of NUPUC BFBT experiments
- 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
No resources found
No enriched resources found
- TRACE VVER-440/V-213 model validation
- Development of a coupled TRACE/PARCS model for KKL and benchmark against the turbine trip test
- Assessment of TRACE V5.0.Patch 4 code against PWR PACTEL loop seal clearing experiment
- Core exit temperature response during an SBLOCA event in the Ascó NPP
- Evaluation of TRACE spacer grid model with FLECHT-SEASET reflood test
- Semiscale S-NC-02 and S-NC-03 natural circulation tests performed by RELAP5/MOD3.3 Patch05
- Feedwater line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- Fuel rod behavior and uncertainty analysis by FRAPTRAN/TRACE/DAKOTA code in maanshan LBLOCA
- Analysis with TRACE code of ROSA test 1.1 : ECCS water injection under natural circulation condition
- Using SNAP/RADTRAD and HABIT to establish the analysis methodology for Maanshan PWR
- IBLOCA analysis for Vandellós-NPP using RELAP5/MOD3.3 sensitivity calculations to EOP actions
- Thermal hydraulic and fuel rod mechanical combination analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD3.3/FRAPTRAN/Python in SNAP interface
- 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
- Assessment of TRACE 5.0 against ROSA-2 test 3 countertest to PKL
- TRACE/RELAP5 comparative calculations for double-ended LBLOCA and SBO
- Loss of flow analysis of Maanshan Nuclear Power Plant with RELAP5/SNAP
- Analysis with TRACE code of Rosa test 1.2 : small LOCA in the hot-leg with HPI and accumulator actuation
- Main stream line break analysis for lungmen ABWR
- Post-test analysis of ROSA-2 test 2 (IBLOCA) with TRACE
- Simulation of ROSA-2 test-2 experiment : application to nuclear power plant
- TRACE/SNAP model establishment of Chinshan nuclear power plant for ultimate response guideline
- Post-test analysis of cold leg small break 4.1% at PSB-VVER facility using TRACE V5.0
- Using TRACE, MELCOR, CFD, and FRAPTRAN to establish the analysis methodology for Chinshan Nuclear Power Plant spent fuel pool
- Post-test calculation of the PKL-2 test G7.1 using RELAP5/MOD3.3
- Simulation of the LSTF-PKL Counterpart G7.1 Test at PKL facility using TRACE 5
- Post-test calculation of the ROSA-2 test 3 using RELAP5/MOD3.3
- Assessment of the wall film condensation model with non-condensable gas in RELAP5 and TRACE for vertical tube and plate geometries
- Simulation of the PKL-G7.1 experiment in a Westinghouse nuclear power plant using RELAP5/MOD3.3
- Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code
- Assessment of critical subcooled flow through cracks in large and small pipes using TRACE and RELAP5
- Uncertainty and sensitivity investigations with TRACE-SUSA and TRACE-DAKOTA by means of post-test calculations of NUPUC BFBT experiments
- RELAP5 analysis of mitigation strategy for extended blackout power condition in PWR
- Steam line break analysis using RELAP5/MOD3.3 for steam generator blowdown load assessment
- RELAP5 and TRACE calculations of LOCA in PWR
- RELAP5 model of a CANDU-6 (Embalse) nuclear power plant : application to a turbine trip event
- The analysis and study of ELAP event and mitigation strategies using TRACE code for Maanshan PWR
- TRACE VVER-1000/V-320 model validation
- RELAP5/MOD3.3 analysis of event with actuation of safety injection system at the loss of external power
- Validation of RELAP5 model of ringhals 4 against a load step test at uprated power
- RELAP5/MOD3.3 model assessment of Maanshan Nuclear Power Plant with SNAP interface
Sponsoring body of
No resources found
No enriched resources found
- Proposal for the development and implementation of an uncertainty and sensitivity analysis module in SNAP
- Assessment against ACHILLES reflood experiment with TRACE V5.0 Patch3
- Post-test thermal-hydraulic analysis of PKL tests F1.1 and F1.2
- TRACE analysis on heat removal decrease accidents for AP1000
- BEPU analysis and benchmark with IIST 2% SBLOCA experiment using TRACE/DAKOTA
- Benchmarking of a generic CANDU reactor with PARCS, MCNP and RFSP
- Research reactor 'MARIA' primary cooling loop transient analysis using RELAP5 Mod 3.3
- Comparison of the U.S. NRC PARCS core neutronics simulator against in-core detector measurements for LWR applications
- Simulation of LSTF upper head break (OECD/NEA ROSA test 6.1) with TRACE code : application to a PWR NPP model
- Simulation of LSTF hot leg break (OECD/NEA ROSA-2 test 1) with TRACE code : application to a PWR NPP model
- The alternate mitigation strategies study of Chinshan BWR/4 by using the LOCA and SBO analysis of TRACE
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