2018
DOI: 10.1016/j.anucene.2018.08.031
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Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT

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Cited by 16 publications
(9 citation statements)
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“…Therefore, it is necessary to analyze the CO 2 diffusion between eggs and the external environment. Fluent (Ansys Inc., Canonsburg, PA, USA) is a popular commercial computational fluid dynamics (CFD) code and is commonly used to analyze fluid distribution 18 . Fabbri et al 19 .…”
Section: Introductionmentioning
confidence: 99%
“…Therefore, it is necessary to analyze the CO 2 diffusion between eggs and the external environment. Fluent (Ansys Inc., Canonsburg, PA, USA) is a popular commercial computational fluid dynamics (CFD) code and is commonly used to analyze fluid distribution 18 . Fabbri et al 19 .…”
Section: Introductionmentioning
confidence: 99%
“…Research reactors are widely used to produce high neutron flux for research, training, education, and irradiation test (Gong et al, 2015;Guo et al, 2018). The fuel assembly used in the research reactors is generally the plate-type fuel assembly, including several fuel plates, two side plates and narrow rectangular channels formed between the fuel plates.…”
Section: Introductionmentioning
confidence: 99%
“…However, fewer thermal-hydraulic investigations of the plate-type fuel assembly have taken the subcooled flow boiling into consideration. Guo et al (2018) conducted the thermal-hydraulic analysis of flow blockage in the fuel assembly in the JRR-3M research reactor. The 3dimensional model of the fuel assembly was built and the flow and heat transfer characteristics were simulated using FLUENT.…”
Section: Introductionmentioning
confidence: 99%
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“…The importance of LOFA and RIA analysis are for safety precaution during operation of research reactor as well as nuclear power plant. Several researches related to flow analysis, LOFA or reactivity in research reactor and nuclear power plant that supporting to this research such as preliminary accident analysis for a conceptual design a 10 MW research reactor [12], dynamic analysis for conceptual core design [14], thermal hydraulic analysis improvement for IEA-R1 reactor [15], characterization of oxide fuel element of RSG-GAS [16], accident safety analysis in JRR-3M [17], analysis of temperature effect on control rod worth in TRR [18], and transient analysis in a downward cooling pool-type material testing reactor [19] reactivity feedback effect on LOFA in PWR [20] and validation using standard code on reactor transient condition [21].…”
Section: Introduction *mentioning
confidence: 99%