889 research outputs found
Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination USi-FeCrAl
Neutronic performance is investigated for a potential accident tolerant fuel
(ATF),which consists of USi fuel and FeCrAl cladding. In comparison
with current UO-Zr system, FeCrAl has a better oxidation resistance but a
larger thermal neutron absorption cross section. USi has a higher
thermal conductivity and a higher uranium density, which can compensate the
reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible
USi-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, USi-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
U, U, and 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 of 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, Np and five MAs (Np, Am, Am,
Cm and 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, Np and Am. In addition, the
inclusion of Np in MOX has no large influence for other MAs, while the
transmutation efficiency of 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
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