Detailed design of242mAm breeding in pressurized water reactors

There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

A high performance oxygen storage material for chemical looping processes with CO2 capture

Chemical looping combustion (CLC) is a novel combustion technology that involves cyclic reduction and oxidation of oxygen storage materials to provide oxygen for the combustion of fuels to CO2 and H2O, whilst giving a pure stream of CO2 suitable for sequestration or utilisation. Here, we report a method for preparing of oxygen storage materials from layered double hydroxides (LDHs) precursors and demonstrate their applications in the CLC process. The LDHs precursor enables homogeneous mixing of elements at the molecular level, giving a high degree of dispersion and high-loading of active metal oxide in the support after calcination. Using a Cu-Al LDH precursor as a prototype, we demonstrate that rational design of oxygen storage materials by material chemistry significantly improved the reactivity and stability in the high temperature redox cycles. We discovered that the presence of sodium-containing species were effective in inhibiting the formation of copper aluminates (CuAl2O4 or CuAlO 2) and stabilising the copper phase in an amorphous support over multiple redox cycles. A representative nanostructured Cu-based oxygen storage material derived from the LDH precursor showed stable gaseous O2 release capacity (∼5 wt%), stable oxygen storage capacity (∼12 wt%), and stable reaction rates during reversible phase changes between CuO-Cu 2O-Cu at high temperatures (800-1000 °C). We anticipate that the strategy can be extended to manufacture a variety of metal oxide composites for applications in novel high temperature looping cycles for clean energy production and CO2 capture. © The Royal Society of Chemistry 2013.

Ultra-thin 242mAm fuel elements in nuclear reactors

There is a growing interest in using 242mAm as a nuclear fuel. The advantages of 242mAm as a nuclear fuel derive from the fact that 242mAm has the highest thermal fission cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal fission is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these fission products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized fission products. 1. Using the fission products themselves for ionic propulsion. 2. Using the fission products in an MHD generator, in order to obtain electricity directly. 3. Using the fission products to heat a gas up to a high temperature for propulsion purposes. In this work, we are not dealing with a specific reactor design, but only calculating the minimal fuel elements' thickness and the energy of the fission products emerging from these fuel elements. It was found that it is possible to design a nuclear reactor with a fuel element of less than 1 μm of 242mAm. In such a fuel element, 90% of the fission products' energy can escape.