Finite Volume Method Based Neutronic Solvers for Steady and Transient Analysis of Molten Salt Reactors
- School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an, 710049 (China)
In recent years, the coupled neutronics and thermal hydraulics analysis becomes more and more important, because it helps to identify the most relevant safety issues without conservative assumptions. In addition, such coupled simulations are essential for the generation IV advanced nuclear power systems such as molten salt reactor (MSR), because the presence of strong coupling between different phenomena, like the transport of delayed neutron precursors by the fluid flow?. Due to the different discretization methods used in reactor physics analysis and thermal-hydraulics analysis, it is hard to couple different kinds of codes. This work presents an attempt to solve the multigroup neutron diffusion equation with the finite volume method which is quite widely used for the numerical simulation of a variety of applications involving fluid flow and heat and mass transfer. The unstructured finite volume method (FVM) based on open source C++ library OpenFOAM was chosen as the development platform?. The objective of this paper is to concisely present the main features of the developed codes in terms of the theoretical background and numerical approach and as can be seen in the following sections, the calculation results obtained by the NDSFoam which solves the steady-state neutron diffusion equation and NDTFoam which solves the transient-state neutron diffusion equation are discussed and compared to the benchmark problems. In this paper, two new neutronic solvers using finite volume method have been developed to perform the neutron physics analysis. The solvers are based on OpenFOAM, and have been tested with several steady state and transient-state benchmark problems. By verifications of the codes, it is proved that the development of NDSFoam and NDTFoam are correct. In our future work, both of these two solvers will be coupled with the precompiled CFD solvers in OpenFOAM to analyze the molten salt fast reactors. (authors)
- OSTI ID:
- 22991999
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society, New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 4 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICAL METHODS AND COMPUTING
COMPUTERIZED SIMULATION
DELAYED NEUTRON PRECURSORS
FAST REACTORS
FLUID FLOW
MOLTEN SALT REACTORS
MOLTEN SALTS
NEUTRON DIFFUSION EQUATION
NEUTRON PHYSICS
NUCLEAR POWER
POWER SYSTEMS
REACTOR PHYSICS
STEADY-STATE CONDITIONS
STRONG-COUPLING MODEL
THERMAL HYDRAULICS