Transient Analysis of Small Molten Salt Reactor : interactions between fission reaction and fuel salt flow in the case of blockage accident
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This paper performed the transient core analysis of a small molten salt reactor in the case of blockage accident. The emphasis is that the numerical model developed in this paper takes into account the interaction between fission reaction and fuel salt flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis in the case of blockade accident are as follows ; (1) neutron multiplication factor hardly changes, (2) the outlet temperature of fuel salt decreases 10 K, and (3) fuel salt and graphite temperatures largely increase at the blockage point, but lower than fuel salt boiling temperature and the molten temperature of the reactor vessel.Keywords:
Molten salt reactor
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The chemistry group within the U.S. Department of Energy Molten Salt Reactor Campaign is developing a model for the molten salt reactor system for assessing the suitability of various used fuel processing schemes. In the long run, the developed model is expected to be a tool for better understanding the interplay between the components of the molten salt reactor system including uranium resources, the molten salt reactor, the used fuel processor (e.g., reprocessing plant), and waste management. In this regard, the report proposes a theoretical framework for integrating the various components of the molten salt reactor system. Rather than being buried in implementation details for seemingly heterogeneous components of the reactor system asking for various modeling techniques, this report proposes to approximate and integrate all components with a unified mathematical framework, a discrete-time linear system. An abstracted molten salt reactor system is given as an example to illustrate the effectiveness of the proposed theoretical framework. A fully developed molten salt reactor system model (based on the proposed framework) is expected to be beneficial for various objectives including the transient and equilibrium analysis of fuel cycle scenarios and the identification of potential technology bottlenecks.
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Electrochemical methods for the separation of fission products from fission material in molten fluoride salt media have been studied in the context of their application within the framework of the developed Molten Salt Reactor fuel cycle. The separation possibilities of selected actinides (U, Th) and lanthanides (Nd, Eu, Gd) in molten eutectic LiF-NaF-KF at 530°C were evaluated by means of cyclic voltammetry. The applicability of different electrochemical techniques is discussed with reference to the new results from this study, and a basic flow sheet for the Molten Salt Reactor fuel cycle is outlined.
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