Abstract: One of the main characteristics of Heavy Water Moderated Reactors is their high production of plutonium. This article demonstrates the possibility of reduction of plutonium and other actinides in Heavy Water Research Reactor. Among the many ways for reducing plutonium production in a heavy water reactor, in this research, changing the fuel from natural Uranium fuel to Thorium-Uranium mixed fuel was focused. The main fissile nucleus in Thorium-Uranium fuels is U-233 which would be produced after neutron absorption by Th-232, so the Thorium-Uranium fuels have some known advantages compared to the Uranium fuels. Due to this fact, four Thorium-Uranium fuels with different compositions ratios were chosen in our simulations; a) 10% UO2-90% THO2 (enriched= 20%); b) 15% UO2-85% THO2 (enriched= 10%); c) 30% UO2-70% THO2 (enriched= 5%); d) 35% UO2-65% THO2 (enriched= 3.7%). The natural Uranium Oxide (UO2) is considered as the reference fuel, in other words all of the calculated data are compared with the related data from Uranium fuel. Neutronic parameters were calculated and used as the comparison parameters. All calculations were performed by Monte Carol (MCNPX2.6) steady state reaction rate calculation linked to a deterministic depletion calculation (CINDER90). The obtained computational data showed that Thorium-Uranium fuels with four different fissile compositions ratios can satisfy the safety and operating requirements for Heavy Water Research Reactor. Furthermore, Thorium-Uranium fuels have a very good proliferation resistance and consume less fissile material than uranium fuels at the same reactor operation time. Using mixed Thorium-Uranium fuels reduced the long-lived α emitter, high radiotoxic wastes and the radio toxicity level of spent fuel.
Abstract: Currently, thorium fuel has been especially noticed
because of its proliferation resistance than long half-life alpha emitter
minor actinides, breeding capability in fast and thermal neutron flux
and mono-isotopic naturally abundant. In recent years, efficiency of
minor actinide burning up in PWRs has been investigated. Hence, a
minor actinide-contained thorium based fuel matrix can confront both
proliferation resistance and nuclear waste depletion aims. In the
present work, minor actinide depletion rate in a CANDU-type nuclear
core modeled using MCNP code has been investigated. The obtained
effects of minor actinide load as mixture of thorium fuel matrix on
the core neutronics has been studied with comparing presence and
non-presence of minor actinide component in the fuel matrix.
Depletion rate of minor actinides in the MA-contained fuel has been
calculated using different power loads. According to the obtained
computational data, minor actinide loading in the modeled core
results in more negative reactivity coefficients. The MA-contained
fuel achieves less radial peaking factor in the modeled core. The
obtained computational results showed 140 kg of 464 kg initial load
of minor actinide has been depleted in during a 6-year burn up in 10
MW power.
Abstract: The presented work is motivated by a French law
regarding nuclear waste management. A new conceptual Accelerator
Driven System (ADS) designed for the Minor Actinides (MA)
transmutation has been assessed by numerical simulation. The
MUltiple Spallation Target (MUST) ADS combines high thermal power (up to 1.4 GWth) and high specific power. A 30 mA and 1
GeV proton beam is divided into three secondary beams transmitted on three liquid lead-bismuth spallation targets. Neutron and thermalhydraulic
simulations have been performed with the code MURE, based on the Monte-Carlo transport code MCNPX. A methodology has been developed to define characteristic of the MUST ADS concept according to a specific transmutation scenario. The reference
scenario is based on a MA flux (neptunium, americium and curium)
providing from European Fast Reactor (EPR) and a plutonium multireprocessing
strategy is accounted for. The MUST ADS reference
concept is a sodium cooled fast reactor. The MA fuel at equilibrium is mixed with MgO inert matrix to limit the core reactivity and
improve the fuel thermal conductivity. The fuel is irradiated over five
years. Five years of cooling and two years for the fuel fabrication are
taken into account. The MUST ADS reference concept burns about 50% of the initial MA inventory during a complete cycle. In term of
mass, up to 570 kg/year are transmuted in one concept. The methodology to design the MUST ADS and to calculate fuel
composition at equilibrium is precisely described in the paper. A detailed fuel evolution analysis is performed and the reference scenario is compared to a scenario where only americium transmutation is performed.
Abstract: The mixed oxide nuclear fuel (MOX) of U and Pu contains several percent of fission products and minor actinides, such as neptunium, americium and curium. It is important to determine accurately the decay heat from Curium isotopes as they contribute significantly in the MOX fuel. This heat generation can cause samples to melt very quickly if excessive quantities of curium are present. In the present paper, we introduce a new approach that can predict the decay heat from curium isotopes. This work is a part of the project funded by King Abdulaziz City of Science and Technology (KASCT), Long-Term Comprehensive National Plan for Science, Technology and Innovations, and take place in King Abdulaziz University (KAU), Saudi Arabia. The approach is based on the numerical solution of coupled linear differential equations that describe decays and buildups of many nuclides to calculate the decay heat produced after shutdown. Results show the consistency and reliability of the approach applied.