Feasibility Study of Thorium-Plutonium Mixed Oxide Assembly In Light Water Reactors

Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light water reactors in the world. The main objective of this paper is not only to show the benefits of using the thorium, but mainly to study how the way thorium is introduced in the fuel affects the neutron parameters. Among these benefits is the possibility of extending the operating cycle length and the reduction of the increasing stockpiles of plutonium. The first investigated method is introducing thorium as (Th,Pu)OX. The second one is a homogeneous model of thorium plutonium oxide. It is carried out by adding an amount of plutonium separated from the uranium oxide cycle at 50 GWd/ton of heavy metal to the same amount of thorium. Thus, we studied three assemblies; the reference assembly is uranium oxide of 4.2% enrichment containing borated water as a moderator of concentration 500 ppm (part per million) of B-10. The second is a (Th,Pu)OX and the third one is an assembly with homogenized thorium plutonium. All three assemblies are modeled using MCNPX. A comparison is held between the results of the three lattices. The factors compared are the effective multiplication factor, the inventory of plutonium and uranium isotopes, and the depletion of B-10, the pin by pin power distribution at 0 and 60 GWd/ton and the relative pin radial power for the three lattices. The comparison is aimed to show the effect on the cycle length, the reduction in the Pu content and the power flattening across the assembly. It is found that the evolution of the multiplication factors shows a similar behaviour using (Th-Pu)OX fuel in the assembly as UOX cycle inspite of lowering the K-eff of fresh (Th-Pu)OX fuel (1.19847). The power flattening which is favorable in reactor operation is clearer in (Th,Pu)OX fuel. It is noticed that the mass of Pu-239 decreases by 1.07% from its initial value at the end of life. For homogeneous (Th,Pu)OX, the mass decreases by 0.0832%. The high power peaking factor for (Th,Pu)OX is not expected to cause significant effects during reactor operation but it can be reduced by adding burnable poisons.

Power Flattening on Design Study of Small Long-Life Boiling Water Reactor (BWR) with Tight Lattice Thorium Nitride Fuel

Preliminary study of thorium based fuel utilization with the addition of Pa-231 on tight lattice boiling water reactor (BWR) has been performed. In previous studies, the use of fuel composition Th-232 and U-233 as well as the use of protactinium as burnable poisons with hexagonal tight lattice fuel cell geometry has resulted the reactor life time of 30 years without refueling [1]. In this study, power flattening has been conducted on the reactor core by using radially heterogeneous fuel. Addition of the Pa-231 is expected to extend lifetime of the BWR core By optimizing the composition of the fuel elements (Th and Pa) at low moderation conditions (tight lattice) it can be obtained the reactor core which can be opeprated over 30 years without refueling or fuel shuffling. The Reactor core has a volume of 17,635.8 liter, power of 620 MWt, operating life of 30 and a maximum excess reactivity value of 0.384% dk/k, could be achieved by using a composition of U-233 enrichment of 8.1 to 11% and the addition of Pa-231 as much as 6.16 to 11.13% with a power density of 35.2 watts/cc.

Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.