The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10⁻⁴); the splitting method could reach a precision of 10⁻² level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors.

The molten chloride salt fast reactor (MCFR) with a closed Th–U fuel cycle is receiving more and more attention due to its excellent performance, such as high solubility of actinides, superior breeding capacity, and good inherent safety. In this work, the neutronics performances for different minor actinides (MA) loadings and operation modes are analyzed and discussed based on an optimized MCFR. The results indicate that online continuous reprocessing can significantly increase the transmutation performance of MAs. In addition, MA loadings have an obvious effect on the neutronics characteristics of the MCFR, and it is helpful for improving the MA transmutation capability and ²³³U breeding performance, simultaneously. When MA = 5 mol%, the average annual MA transmutation mass and incineration mass can achieve about 53 kg and 13 kg, respectively, and the corresponding annual net production of ²³³U is 250 kg. When MA = 33.5 mol%, the annual MA transmutation mass and incineration mass can be about 310 kg and 77 kg, respectively, and the corresponding annual net production of ²³³U is 349 kg. However, when the MA loadings exceed 10%, the corresponding k_eff will exceed 1.1 for decades, even if only Th is continuously fed online. The results also indicate that the transmutation ratio (TR) and incineration ratio (IR) of MA increase and reach maximum values in the first decades for all the different MA loadings, which means MA may be fed into the fuel salt to improve its transmutation capability. Moreover, though MA loading will increase the level of radiotoxicity of the core in the early stage of burnup, the radiotoxicity of MA will drop rapidly after a brief rise during the operation. It can also be found that the temperature coefficient of reactivity (TCR) of all different MA loadings can be negative enough to maintain the safety of the MCFR during the whole operation, although it decreases in the beginning of life (BOL) with the increasing MA loading. Furthermore, the evolution of an effective delayed neutron fraction (EDNF) is also researched and discussed, and the EDNF varies most significantly when loading MA = 35.5 mol%, with a range of 273 to 310 pcm over the entire 100 years of operation.

Long-lived minor actinides (MAs) are one of the primary contributors to the long-term radiological hazards of nuclear waste, and the buildup of MAs is hampering the development of nuclear power. The transmutation of MAs in reactors is regarded as a potential way to replace direct disposal to reduce the impact of MA on the environment and improve the utilization of fuel. Due to its superior features, such as outstanding neutron economy, no fuel assembly fabrication, high neutron flux, and especially online refueling and reprocessing, the molten chloride salt fast reactor (MCFR) is regarded as one of the potential reactors for MA incineration. In this work, MA transmutation capability and ²³³U breeding performance for an optimized MCFR have been evaluated in different scenarios. The results show that the MA transmutation capability and ²³³U breeding performance with online transuranic elements (TRU) and ²³²Th feeding scenario are improved significantly compared with the case in online ²³³U and ²³²Th feeding, when the initial MA loading is 5 mol%, the total mass of MA transmutation and MA incineration is 7160 kg and 1759 kg during the whole 100 years operation under online TRU and ²³²Th feeding scenario, and the corresponding average annual net production of ²³³U is 450 kg, however, the MA transmutation amount, MA incineration amount and average annual net production of ²³³U for online ²³³U and ²³²Th feeding scenario is 5298 kg, 1315 kg, and 249 kg, respectively. In addition, the research also shows that the increase in initial loading of MA has no obvious effect on the improvement of the ²³³U breeding performance but can improve the transmutation efficiency of MA under online TRU and ²³²Th feeding scenarios. Furthermore, if ²³³U is continuously extracted online from the core during the operation, the ²³³U breeding performance will be significantly improved, but it will deteriorate the safety performance, such as the fuel temperature coefficient of reactivity (TCR) and the effective delayed neutron fraction (EDNF), more importantly, it will also put forward higher requirements for the immature online reprocessing technology.

In molten salt reactors (MSRs), a small amount of inert gas could be entrained from the free liquid surface to the primary loop, which may have obvious impacts on the heat transfer performance of the heat exchanger, the reactivity of the core, and the migration of insoluble fission products. It is necessary to understand how the bubbles flow into the reactor core, and what kinds of size distribution should they be. Meanwhile, the heat exchanger is an important and complicated place before the gas enters the core, in which the bubbles may exhibit complex behavior, such as coalescence, breakup, or retention. This research employs a coupling method between the Eulerian two‐phase flow model (ETFM) and the population balance model (PBM) to simulate the two‐phase flow of bubbles entrained in molten salt on the shell side of a vertical U‐tube heat exchanger. The gas fraction and poly‐dispersed bubble size distribution are analyzed under different calculation methods and input conditions. The results show that the distribution of gas volume fraction and bubble size are significantly influenced by the characteristics of the flow field, and the salt flow can also be affected by the bubble distribution. The bubbles exhibit obvious non‐uniformity distribution, especially in the center of the separation and the backflow vortexes, and a significant accumulation of gas attachment occurs behind the baffles. The interfacial area concentration and surface heat transfer coefficient are also discussed with or without the bubble distribution. All indicate that a precise bubble spatial and size distribution is necessary when in the simulation of multiphase flow in MSRs.

In molten salt reactors, multiphase transport of fission products produces nonuniform nuclide inventories throughout the primary loop, thereby affecting subsequent decay heat evolution. Although fission product transport has been widely studied, its influence on post-shutdown wall-deposited decay heat remains insufficiently characterized. Here, the bubble transport capability of the Thorium Fission Products Migration Code (ThorFPMC) was extended using a drift-flux model to investigate the effects of operating void fraction on the magnitude, spatial distribution, nuclide composition, and subsequent evolution of wall-deposited decay heat under an idealized post-drainage condition. Results showed that, for the baseline bubble diameter of 0.508 mm, the initial wall-deposited decay heat decreased from 41.33 to 6.63 kW as the operating void fraction increased, corresponding to an 84.0% reduction; bubble diameter sensitivity calculations preserved the same monotonic trend. The source remained strongly nonuniform, with the heat exchanger consistently dominant, while increasing the void fraction reduced the core fraction and increased the relative external-loop contribution. Nuclide contributions evolved from a broad early distribution to long-term dominance by ⁹⁵Nb and ¹⁰³Ru, with precursor feeding shaping several important daughter nuclides. These results link operating period multiphase transport to component-resolved residual decay heat sources and provide a basis for operating-condition-dependent source terms in shutdown thermal and system analyses.