Results 41 to 50 of about 1,865 (177)
Assessment of RELAP/SCDAPSIM/MOD3.4 Prediction Capability with Severe Fuel Damage Scoping Test
Science and Technology of Nuclear Installations, 2017 The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure during accident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for current ...
Noppawan Rattanadecho +4 more
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RELAP5 Calculations of Bethsy 9.1b Test
Science and Technology of Nuclear Installations, 2012 Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed.
Andrej Prošek
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Loss of offsite power accident analysis in a VVER-1000/V446 nuclear power plant [PDF]
Nuclear Technology and Radiation Protection, 2019 The aim of this study is to present a thermo-hydraulic analysis of the loss of offsite power accident in VVER-1000/V446 nuclear power plant using the RELAP5 code.
Esfandiari Mohsen +3 more
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Heat Transfer Boundary Conditions in the RELAP5-3D Code [PDF]
Volume 3: Thermal Hydraulics; Instrumentation and Controls, 2008 The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c ...
Riemke, Richard A. +2 more
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Advanced Presentation of BETHSY 6.2TC Test Results Calculated by RELAP5 and TRACE
Science and Technology of Nuclear Installations, 2012 Today most software applications come with a graphical user interface, including U.S. Nuclear Regulatory Commission TRAC/RELAP Advanced Computational Engine (TRACE) best-estimate reactor system code.
Andrej Prošek, Ovidiu-Adrian Berar
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RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1 [PDF]
, 1995 The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents ...
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RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation
Science and Technology of Nuclear Installations, 2018 The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in ...
Andrej Prošek, Marko Matkovič
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Alexandria Engineering Journal, 2018 Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-
Z. Tabadar +3 more
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Application of RELAP5/Mod3.3 – Fluent coupling codes to CIRCE-HERO
Journal of Physics: Conference Series, 2019 Abstract This paper presents the work ongoing at the DICI (Dipartimento di Ingegneria Civile e Industriale) of the University of Pisa on the application of coupled methodology between Fluent CFD code and RELAP5/Mod3.3 system code. In particular, this methodology was applied to the LBE-water heat exchanger HERO, with the aim to analyse ...
Forgione N. +6 more
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Nuclear Engineering and Technology, 2018 An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with ...
Takeshi Takeda
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