U.S. Nuclear Regulatory Commission. Division of Systems Analysis
VIAF ID: 180148037 ( Corporate )
Permalink: http://viaf.org/viaf/180148037
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Works
Title | Sources |
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The analysis and study of ELAP event and mitigation strategies using TRACE code for Maanshan PWR | |
Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code | |
Analysis with TRACE code of Rosa test 1.2 : small LOCA in the hot-leg with HPI and accumulator actuation | |
Application to a turbine trip event | |
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 | |
Assessment of TRACE 5.0 against ROSA-2 test 3 counterpart test to PKL | |
Assessment of TRACE V5.0 Patch 4 code against Pressurized Water Reactor PArallel Channel TEst Loop loop seal clearing experiment | |
Cladding behavior during postulated loss-of-coolant accidents | |
Core exit temperature response during an Small Break LOss of Coolant Acicident event in the Ascó Nuclear Power Plant | |
Critical heat flux data used to generate the 2006 Groeneveld lookup tables | |
The development and application of Kuosheng (BWR/6) nuclear power plant TRACE/SNAP model | |
Evaluation of TRACE spacer grid model with full-length emergency core heat transfer separate effects and systems effects test reflood test | |
Intermediate Break Loss Of Coolant Accident analysis for Vandellós Nuclear Power Plant using RELAP5/MOD3.3 sensitivity calculations to Emergency Operaton Procedure actions | |
Investigation of the loop seal clearing phenomena for the Advanced Thermal-hydraulic test Loop for Accident Simulation Direct Vessel Injection line and cold leg Small Break Loss Of Coolant Accident tests using Multi-dimensional Advanced Reactor Safety program-Korea Standard and Reactor Excursion and Leak Analysis Program 5/MOD3.3 | |
Laminar hydraulic analysis of a commercial pressurized water reactor fuel assembly | |
A quantitative impact assessment of hypothetical spent fuel reconfiguration in spent fuel storage casks and transportation packages | |
Reactor Excursion and Leak Analysis Program 5 analysis of mitigation strategy for extended blackout power condition in Pressurized Water Reactor | |
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 | |
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, 2010: | |
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 | |
Technical basis for the containment protection and release reduction rulemaking for boiling water reactors with Mark I and Mark II containments | |
TRACE assessment for effect of spacer grid in RBHT reflood heat transfer experiments | |
TRACE VVER-1000/V-320 model validation | |
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 | |
VARSKIN 6 : a computer code for skin contamination dosimetry |