Zhangpeng Guo

COUPLE, A Time-Dependent Coupled Neutronics and Thermal-Hydraulics Code, and its Application to MSFR

Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.Copyright

The Optimization Design of Lower Plenum and Distribution Plates in MSR

Molten salt reactor was one of six Generation IV reactor types, which uses the liquid molten salt as the coolant and fuel solvent. In transmutation of actinides and long-lived fission products have marked advantages. As a liquid reactor the physical property and thermo-characteristic is different to solid fuel and water coolant reactors, which was influenced by many factors. MOSART was one of the advanced molten salt reactors concept design, which can burners TRU from LWR spent fuel. The reactor core does not contain graphite structure elements, so the flow pattern was potentially complex and may significantly affect the fuel temperature distributions. The optimizations of the salt flow may be needed, the present work designed three core models and three kinds of distribution plates to investigate the influence of lower plenum and distribution plates to thermohydraulics characteristics of the reactor core with CFD method use software FLUENT. Velocity field and maximum temperature of the core was simulated in each model at different mass flow rate.© 2013 ASME

Coupled Neutronics/Thermal-Hydraulics for Analysis of Molten Salt Reactor

The Generation IV international Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. This work was helpful for further design analysis and operation of MSR.Copyright

Numerical Analysis for a Molten Salt Reactor in the Presence of Fissile Lump

Molten salt reactors (MSRs) have seen a marked resurgence of interest over the past few decades, highlighted by their inclusion as one of the six Generation IV reactor types. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this paper, the attention is focused on the behaviors of a MSR in the presence of localized perturbations caused by fissile precipitates. A neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors, and the group constants dependent on the temperature are calculated by the code DRAGON. In addition, the k-epsilon turbulent model is adopted to establish the flow and heat transfer. The thermo-hydraulic and neutronic models which are coupled through the temperature, heat source and velocity are coded in a program. The effects of the localized perturbation on the distributions of power, temperature, neutron fluxes and delayed neutron precursors are obtained and discussed in detail. The results provide some valuable information for the research and design of this new generation reactor.Copyright