Results 31 to 40 of about 362 (143)
Frontiers in Energy Research, 2022 Simulations of radiation heat transfer in fuel rod bundles are necessary for the thermal hydraulic design and safety analysis of open lattice gas-cooled reactors, which always operate at high temperatures.
Qian Sun, Yu Ji, Jun Sun
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Science and Technology of Nuclear Installations, 2015 In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and
Viet-Anh Phung, Pavel Kudinov
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A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3
, 1995 The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for ...
Andreani, M, Analytis, G T, Aksan, S N
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Development and analysis of passive residual heat-removal system on secondary side of small reactor
He jishuBackgroundPassive residual heat-removal system (PRHRS) based on natural circulation has been widely used in small reactors connected on secondary side loop.PurposeThis study aims to develop and analyze a passive residual heat-removal system on the ...
DONG Botong +7 more
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Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1 [PDF]
Nuclear Technology and Radiation Protection, 2011 The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4.
Muhammad Atta +2 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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Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.In the design of nuclear power plant systems, the thermal–hydraulic response in the containment during a loss‐of‐coolant accident (LOCA) plays a crucial role. In this study, the advanced small modular reactor NHR200‐II is selected as the research object.
Yan Wang +3 more
wiley +1 more source
SCDAP/RELAP5 independent peer review [PDF]
, 1993 The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall ...
Corradini, M. L. +7 more
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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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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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