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Descripción
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| This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel. | |
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Internacional
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Si |
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DOI
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10.1088/1742-6596/112/3/032046 |
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Edición del Libro
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0 |
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Editorial del Libro
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ISBN
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1742-6596 |
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Serie
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Título del Libro
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Journal of Physics: Conference Series |
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Desde página
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1 |
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Hasta página
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4 |