The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system
Tutor / director / avaluadorAbram, Timothy
Tipus de documentProjecte/Treball Final de Carrera
Condicions d'accésAccés restringit per decisió de l'autor
The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. Moreover it is also studied the effect of the burnable poisons in the core and the control rods. Different concentrations of poison are considerate and the control rods are studied in different positions and materials. Altogether 140 criticality calculations have been made. The report can be split in two parts. The first efforts are leaded to build the MCNP model that reproduces the actual U-Battery in the program. The 3-D model is a reliable reproduction of the core. Since the description of the small coated particles in the graphite matrix that conform the fuel rods, the assembly with the fuel rods, coolant channels and burnable poisons rods until the construction of the core with the assembly array and reflectors. A good geometry description guarantees reliable keff results. The second part is based in the model. To reproduce each condition in the core to study the behaviour of keff the model is slightly modified and MCNP is run to obtain the results. Once the results are obtained they will be validated against similar analyses conducted at Delft using different code suite. The study would be extended in the future to study the burnup modelling.
MatèriesNuclear batteries, Monte Carlo Method, Neutron transport theory, Bateries nuclears, Monte Carlo, Mètode de, Transport del neutró, Teoria del
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