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Investigations on RELAP5-3D to RELAP5-3D Coupling Methodology by PVMEXEC
Volume 3: Thermal-Hydraulics, 2016In the framework of a BEPU (Best Estimate Plus Uncertainty) approach within the licensing process of a nuclear power plant, the need to extend the resources of nuclear system thermal-hydraulics codes, such as RELAP5-3D, arises to allow more detailed simulations of the complex 3D reality of Nuclear Power Plants (NPPs), either under normal steady-state ...
Valeria Parrinello +3 more
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COBRA/RELAP5: A Merged Version of the COBRA-TF and RELAP5/MOD3 Codes
Nuclear Technology, 1992The best-estimate thermal-hydraulic codes RE-LAP5/MOD3 and COBRA-TF were adopted to the Apollo DN 10000 workstation and subsequently merged.
Sang Yong Lee +3 more
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2016
In this chapter, we consider the full 1D TFM of RELAP5 for bubbly vertical flows and assess its linear stability behavior and material wave propagation capabilities in light of the linear stability analyses of Chap. 5, i.e., the characteristics and the dispersion relation.
Martín López de Bertodano +3 more
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In this chapter, we consider the full 1D TFM of RELAP5 for bubbly vertical flows and assess its linear stability behavior and material wave propagation capabilities in light of the linear stability analyses of Chap. 5, i.e., the characteristics and the dispersion relation.
Martín López de Bertodano +3 more
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Nuclear Technology, 2016
To support the functionality of RELAP5-3D for best-estimate reactor simulation code, a variety of utility programs were developed at Idaho National Laboratory.
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To support the functionality of RELAP5-3D for best-estimate reactor simulation code, a variety of utility programs were developed at Idaho National Laboratory.
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RELAP5 investigation on subchannel flow instability
Kerntechnik, 2016Abstract Two-phase flow instability is a vitally important area of study for a large number of industrial systems. Density Wave Oscillation (DWO) is the most common type of flow instability caused by the change in flow rate or power in boiling systems.
S. Wang, B.-W. Yang, A. Liu, X. Liu
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2018
The paper presents applicability of built-in RELAP5/MOD3.2 cladding deformation model for VVER-1000 fuel with cladding of Zr+1 % Nb alloy. Experimental data and simplified model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the hot channel blockage after cladding swelling and rupture in the ...
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The paper presents applicability of built-in RELAP5/MOD3.2 cladding deformation model for VVER-1000 fuel with cladding of Zr+1 % Nb alloy. Experimental data and simplified model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the hot channel blockage after cladding swelling and rupture in the ...
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The RELAP5-FORCE MOD2 Code, A Hydrodynamic Forcing Function Calculation Version of RELAP5
Nuclear Technology, 1990A version of the RELAP/MOD2 code has been developed to allow the direct calculation of hydrodynamic piping load histories. The code allows the calculation of piping forcing functions to evaluate the transient thermofluid conditions. A description of this code, RELAP5-FORCE MOD2 is provided.
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DF-4 Analysis Using SCDAP/RELAP5
Nuclear Technology, 1992In this paper the SCDAP/RELAPS severe accident analysis computer code, developed at the Idaho National Engineering Laboratory, is used to analyze the fourth in a series of debris formation experiments. The debris formation-four (DF-4) experiment deals with heatup and meltdown of a boiling water reactor (BWR)-representative fuel and control blade ...
Judith K. Hohorst, Chris M. Allison
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Comparison of MYRRHA RELAP5 MOD 3.3 and RELAP5-3D Models on Steady State and PLOF Transient
Nuclear Technology, 2016AbstractThe MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project aims at the construction of a pool-type subcritical accelerator-driven system that could also operate a...
D. Castelliti, T. Hamidouche
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RELAP5-SCDAP RIA Type Tests Model
Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy, 2006Extended tests on fuel behavior in Reactivity Initiated Accidents type were carried out in the Annular Core Pulse Reactor (ACPR). The tests were performed in a specially designed capsule filled with 7 liters of water at atmospheric pressure and at temperature of 25 C.
Gheorghe Negut +2 more
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