Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3_3Si2_2-FeCrAl

Neutronic performance is investigated for a potential accident tolerant fuel (ATF),which consists of U$_3$Si$_2$ fuel and FeCrAl cladding. In comparison with current UO$_2$-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U$_3$Si$_2$ has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U$_3$Si$_2$-FeCrAl fuel-cladding systemis taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly.These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U$_3$Si$_2$-FeCrAl system is a potential ATF candidate from a neutronic view

A simple formula for local burnup based on constant relative reaction rate per nuclei

A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of $^{235}$U, $^{238}$U, and $^{239}$Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results as the MC calculation from the starting to high burnup level, but takes just a few minutes. The relative reaction rates are found to be almost independent on the radius (except $(n,\gamma)$ of $^{238}$U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance on the accuracy and the cost on time

Study of Minor Actinides Transmutation in PWR MOX fuel

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to be an efficient method to treat the long-lived radionuclides in spent fuel. Some Minor Actinides (MAs) have very long half-lives among the radionuclides in the spent fuel. Accordingly, the study of MAs transmutation is a significant work for the post-processing of spent fuel. In the present work, the transmutations in Pressurized Water Reactor (PWR) mixed oxide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs, $^{237}$Np and five MAs ($^{237}$Np, $^{241}$Am, $^{243}$Am, $^{244}$Cm and $^{245}$Cm) are incorporated homogeneously into the MOX fuel assembly. The transmutation of MAs is simulated with different initial MOX concentrations. The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for the two kinds of MAs primarily generated in the UOX fuel, $^{237}$Np and $^{241}$Am. In addition, the inclusion of $^{237}$Np in MOX has no large influence for other MAs, while the transmutation efficiency of $^{237}$Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be a new, efficient nuclear spent fuel management method for the future nuclear power generation