# Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

• The Molten Salt Reactor (MSR), one of the Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition.

•  [1] . Rosenthal M W, Kasten P R, Briggs R B. Nuclear Appli-cation Technology, 1970, 8(2): 1072. Dulla S, Ravetto P, Rostagno M M. Neutron Kinetics of Fluid-Fuel Systems by the Quasi-Static Method. Nuclear Energy, 2004, 31: 17093. Ignatiev V, Feynberg O, Myasnikov A et al. Mosart Fuels and Container Materials Study: Case for Na, Li, Be/F Sol-vent System. In: Proceedings of the 2003 ANS/ENS Inter-national Winter Meeting. New Orleans: American Nuclear Society, 20034. Merzlyakov V, Ignatiev V. Measurement of Transport Properties for Molten Na, Li, Be/F Mixture. In: the11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics. Avignon: American Nuclear Society,20055. Wang S, Rineiski A, Maschek W. Nuclear Engineering and Design, 2006, 236: 15806. Kondo, Sa, Tobita Y, Morita Y et al. Current Status and Validation of the SIMMER-III LMFR Safety Analy-sis Code. In: Proceedings of the ICONE-7. Tokyo: JapanSociety of Mechanical Engineering, 19997. Marleau G, Hebert A, Roy R. A User's Guide for DRAGON3.05C. Canada: Ecole Polytechnique de Montreal, 2006.1-1758. XIE Z S. Numerical Calculation of Nuclear Reactor Neutronics. Beijing: Atomic Energy Press, 1997, 29-31 (in Chinese)

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ZHANG Da-Lin, QIU Sui-Zheng, LIU Chang-Liang and SU Guang-Hui. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt[J]. Chinese Physics C, 2008, 32(8): 624-628. doi: 10.1088/1674-1137/32/8/007
ZHANG Da-Lin, QIU Sui-Zheng, LIU Chang-Liang and SU Guang-Hui. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt[J]. Chinese Physics C, 2008, 32(8): 624-628.
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沈阳化工大学材料科学与工程学院 沈阳 110142

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## Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

###### Corresponding author: ZHANG Da-Lin,

Abstract:

The Molten Salt Reactor (MSR), one of the Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition.

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