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• W3140369145 abstract "Molten Salt Reactors (MSRs) are one of the systems retained by Generation IV as a candidate for the next generation of nuclear reactors. This type of reactor is particularly well adapted to the thorium fuel cycle (Th- {sup 233}U) which has the advantage of producing less minor actinides than the uranium-plutonium fuel cycle ({sup 238}U- {sup 239}Pu). In the frame of a major re-evaluation of the MSR concept and concentrating on some major constraints such as feasibility, breeding capability and, above all, safety, we have considered a particular reactor configuration that we call the 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum. This reactor is presented in the first section. MSRs benefit from several specific advantages which are listed in a second part of this work. Beyond these advantages of the MSR, the level of the deterministic safety in such a reactor has to be assessed precisely. In a third section, we first draw up a list of the reactivity margins in our reactor configuration. We then define and quantify the parameters characterizing the deterministic safety of any reactor: the fraction of delayed neutrons, and the system's feedback coefficients that are here negative. Finally, using a simple point-kinetic evaluation, we analyze how these safety parameters impact the system when the total reactivity margins are introduced in the MSR. The results of this last study are discussed, emphasizing the satisfactory behavior of the MSR and the excellent level of deterministic safety which can be achieved. This work is based on the coupling of a neutron transport code called MCNP with a materials evolution code. The former calculates the neutron flux and the reaction rates in all the cells while the latter solves the Bateman equations for the evolution of the materials composition within the cells. These calculations take into account the input parameters (power released, criticality level, chemistry, etc.), by adjusting the neutron flux or the materials composition of the core on a regular basis. Our calculations are based on a precise description of the geometry and consider several hundreds of nuclei with their interactions and radioactive decay; they allow a thorough interpretation of the results. All the data discussed in this paper result from the evolution of the reactor over 100 years. (authors)" @default.
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• W3140369145 date "2006-07-01" @default.
• W3140369145 modified "2023-09-24" @default.
• W3140369145 title "Molten salt reactor: Deterministic safety evaluation" @default.
• W3140369145 hasPublicationYear "2006" @default.
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