Neutronic Analysis of Flux Dispersion in a Multi-Layered, (D-T) Driven Hybrid Blanket


Abstract: The concept of `hybrid blanket' is based on the placement of the nuclear fuel layer, which is a fertile material and fissionable by the fusion neutrons, at the front or the rear sides of the tritium breeding zone so that, in addition to gaining fission energy, a fissile fuel is produced. The neutronic flux distribution (neutron spectrum) along the radial direction varies with the type of material and the geometry used in the blanket, and also according to whether it is multi-layered or single-layered. The flatness of neutron flux is important for the flatness of power production. In this study, a cylindrical hybrid blanket with a mixture fuel (UO$_2$+CmO$_2$), C reflector, and LiO$_2$ tritium breeding material is neutronically analyzed. While the hybrid blanket is analyzed by separating it into 79 intervals, the neutronic flux distribution in the fuel and other layers are calculated by transforming the reflector and the tritium breeding zones to one-, two-, and three-layered structures respectively according to their volumes. The multi-layered structure and the partial moderation both lead to neutron economy and smoother flux distribution. Therefore, it can be concluded that the multi-layered blanket structure is advantageous for a flat power output. The flux functions of the fuel and other zones are calculated by using the transport code ANISN-ORNL and the data libraries DLC-36 and SINEX. The results are presented graphically as a function of the radius of the reactor, and compared with each other.

Keywords: Hybrid Blanket, Fission, Fusion, Multi-Layer Blanket, Fissile and Fusile Breeding.

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