Results 11 to 20 of about 326 (121)
Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor [PDF]
Nuclear Technology and Radiation Protection, 2006 RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants.
Bousbia-Salah Anis +5 more
doaj +1 more source
Development and Assessment of the COBRA/RELAP5 Code. [PDF]
Journal of Nuclear Science and Technology, 1997 The COBRA/RELAP5 code has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3.2, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This paper provides the integration scheme of the two codes. The results of developmental assesments are also provided,
JEONG, Jae-Jun +2 more
openaire +1 more source
Validation of RELAP5 MOD3.3 code for Hybrid-SIT against SET and IET experimental data
Nuclear Engineering and Technology, 2020 We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with ...
Ho Joon Yoon +3 more
doaj +1 more source
Science and Technology of Nuclear Installations, 2017 Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions.
Eltayeb Yousif +3 more
doaj +1 more source
Brayton cycle numerical modeling using the RELAP5-3D code, version 4.3.4
Brazilian Journal of Radiation Sciences, 2019 This work contributes to enable and develop technologies to mount fast microreactors, to generate heat and electric energy, for the purpose to warm and to supply electrically spacecraft equipment and, also, the production of nuclear space propulsion ...
Eduardo Pedroso Longhini +4 more
doaj +1 more source
Yuanzineng kexue jishu, 2023 Compared with traditional large scale pressurized water reactor, the volume of the containment of small pressurized water reactor is small. During a loss of coolant accident (LOCA), the containment pressure of small pressurized water reactor rises ...
QIN Bitan;ZHAO Dihong;HONG Gang;LU Jiaming;ZHANG Yaoli
doaj
Transient analysis of OSU-MASLWR with RELAP5
Journal of Physics: Conference Series, 2022 Abstract The present paper deals with the assessment of the original and a modified version of RELAP5/MOD3.3 against the OSU Multi Application Small Light Water Reactor (OSU-MASLWR). The new implemented features regard suitable correlations for the heat transfer coefficient evaluation in helical geometry.
Molinari M. +3 more
openaire +1 more source
Development and Verification of Visual Interactive Simulation Platform for Reactor System
Yuanzineng kexue jishu, 2023 RELAP5 is one of the most commonly used programs for reactor thermal hydraulic analysis. Due to the early developing time, RELAP5 does not have a friendly graphical user interface and interactive interface, resulting in the lack of human-computer ...
ZHANG Penghui;LU Guoqing;CHEN Ronghua;XU Wanqing;SU Guanghui;TIAN Wenxi;QIU Suizheng
doaj
He jishu, 2022 BackgroundThe natural circulation chloride cooled fast reactor (N3CFR) has the characteristics of simple structure, high inherent safety and good economy. It is an advanced fourth-generation nuclear energy system with development potential.
SONG Shiyang +3 more
doaj +1 more source
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
doaj +1 more source