Research of Neutron-thermal Hydraulic Coupling Calculation Method in Fast Reactor Code System MOSASAUR [PDF]
Yuanzineng kexue jishuThe nuclear reactor core is a highly heterogeneous system where various physical phenomena are interrelated and coupled. The complex coupled interaction necessitates multi-physics calculations to realize more accurate and realistic simulations in core ...
ZHANG Bin, WANG Lianjie, LOU Lei, ZHAO Chen
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Research on Thermal-hydraulic Calculation Method of Steam Generator for Integrated Fast Reactor [PDF]
Yuanzineng kexue jishuIntegrated fast reactors emerge as the future direction for sodium cooled fast reactor systems. Steam generator is critical components in reactor systems. Their design directly impacts integrated fast reactor construction quality.
YU Qi, ZHU Lina, ZHU Huanjun, HOU Bin
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TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4
International Journal of Technology, 2017 Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes.
Anhar Riza Antariksawan +4 more
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You-qi chuyunObjectiveThe transient operation simulation technology for natural gas pipeline networks is essential for peak shaving, economical and safe operation, and diagnosing abnormal service conditions.
Chunyu YU +7 more
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Analysis of heat balance of piston hydraulic damper system installed on thrust bearing
Zhongguo Jianchuan Yanjiu, 2018 [Objectives] This paper aims to analyze the thermal equilibrium performance of a piston hydraulic damper in a stable working state so as to solve the problem in which the oil circuit of the hydraulic damping system of a longitudinal vibration-reduction ...
CHEN Fan +3 more
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SUBSALS: a subchannel thermal-hydraulic code for IRT type fuel analysis
Acta Polytechnica CTU Proceedings, 2022 The LVR-15 research reactor is operated with a tube type fuel IRT-4M. Due to fuel’s unique concentric square annular shape with, coolant flow is subject to significant pressure driven crossflow.
Tomáš Adámek
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Increasing the thermal efficiency of water heating boilers by improving design parameters [PDF]
E3S Web of Conferences, 2023 Currently, research is being conducted all over the world to create clearly defined operating, technological and design parameters that ensure the continuity of hydrodynamic and thermal processes, which will improve the energy efficiency of heat supply ...
Arifjanov Aybek, Kurbonov Kozim
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Development and validation of reactor nuclear design code CORCA-3D
Nuclear Engineering and Technology, 2019 The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback,
Ping An +5 more
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Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor [PDF]
Nuclear Technology and Radiation Protection, 2022 The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA.
Hutli Ezddin, Kridan Ramadan
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Problems of calculation of heat transfer crisis in fuel assembles of water cooled reactors [PDF]
Ядерна фізика та енергетика, 2018 Current problem of the ensuring reliability of the results of mathematical computer simulation of the operational modes of water-cooled nuclear reactors is considered in this article.
G. I. Sharaevsky +3 more
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