A code system has been developed in this paper for the dynamics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY and TANSY_K which is based on OpenFOAM to perform the steady-state full-core coupled simulations and dynamics simulation. For verification and application of the codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. For the further study of the characteristics of MSRs, several transients like the code-slug transient, unprotected loss of flow transient and overcooling transient have been analyzed. The numerical results indicated that the TANSY and TANSY_K codes with the cross section library generated by OpenMC has the capability for the dynamics analysis of MSRs.
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2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5781-6
PROCEEDINGS PAPER
Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors Available to Purchase
Tianliang Hu,
Tianliang Hu
Xi’an Jiaotong University, Xi’an, China
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Liangzhi Cao,
Liangzhi Cao
Xi’an Jiaotong University, Xi’an, China
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Hongchun Wu,
Hongchun Wu
Xi’an Jiaotong University, Xi’an, China
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Kun Zhuang
Kun Zhuang
Xi’an Jiaotong University, Xi’an, China
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Tianliang Hu
Xi’an Jiaotong University, Xi’an, China
Liangzhi Cao
Xi’an Jiaotong University, Xi’an, China
Hongchun Wu
Xi’an Jiaotong University, Xi’an, China
Kun Zhuang
Xi’an Jiaotong University, Xi’an, China
Paper No:
ICONE25-67316, V003T02A051; 7 pages
Published Online:
October 17, 2017
Citation
Hu, T, Cao, L, Wu, H, & Zhuang, K. "Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application. Shanghai, China. July 2–6, 2017. V003T02A051. ASME. https://doi.org/10.1115/ICONE25-67316
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