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Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

Klara Insulander Björk (Institutionen för teknisk fysik, Nukleär teknik) ; Valentin Fhager ; Christophe Demazière (Institutionen för teknisk fysik, Nukleär teknik)
Progress in Nuclear Energy (0149-1970). Vol. 53 (2011), 6, p. 618-625.
[Artikel, refereegranskad vetenskaplig]

With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel.

Nyckelord: BWR; Neutronics; Thorium



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Denna post skapades 2011-08-18. Senast ändrad 2015-09-01.
CPL Pubid: 144561

 

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Institutionen för teknisk fysik, Nukleär teknik (2006-2015)

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