The oxygen potentials at 1273 K of mixed oxide (MOX) fuels with Am and 26 kinds of fission product elements (FPs), simulating low-decontaminated MOX fuel and high burn-up of up to 250 GWd/t, have been measured by using thermogravimetric analysis (TGA). The oxygen potentials for simulated low-decontaminated MOX fuels were higher than the fuels without FPs and increased with increasing simulated burn-up.

In order to evaluate the radiation response of 12YWT nanostructured ferritic steel to high dose neutron irradiation, the solute distribution, and size, number density, and compositions of nanoclusters in the unirradiated condition and after neutron irradiation to a dose of 3 dpa at a controlled temperature of 600 °C were estimated by atom probe tomography. No statistical difference in the average size or size distribution of the nanoclusters was found between the unirradiated and irradiated conditions. Therefore, these nanostructured ferritic steels are promising candidate materials for use under extreme conditions in future generations of advanced reactors.

FF-GDMS and MIC-ICP-MS methods were developed for the determination of mg/kg- and μg/kg-level B, Mg, Al, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Sb, W and Pb in nuclear-grade graphite. Consistent results have been achieved in determining trace elements like B, Ti, Cr, Mn, Zr, Sb and Pb by both methods, which vary mostly less than ±30%, and are in line with the manufacturer reference values. On Mg, Al, Fe, Co, Zn, Mo and W, FF-GDMS analyses also show good agreement with the manufacturer's data. Continuing efforts in identifying source of interference, which has limited the MIC-ICP-MS analysis of these elements, is currently underway.

We proposed a new concept for densification of inert matrix fuels containing minor actinides. In this concept, magnesium silicates which are both a naturally-occurring material and asbestos waste were used as a sintering additive which protects public health by safely disposing of the asbestos waste. In this study, the effects of magnesium silicate additives on the densification behaviors of MgO, Mo and CeO2 were experimentally investigated. The densities of MgO and CeO2 pellets increased with only 1 wt.% additives of MgSiO3 and Mg2SiO4. The densities of Mo pellets showed little change with additives.

A 1 μg High Burn Up Structure (HBS) fragment was extracted from a UO2 fuel pellet irradiated for 7 cycles in a EDF Pressurised Water Reactor (PWR). In situ examinations were performed with an Environmental Scanning Electron Microscope (ESEM) in order to characterize UO2 microstructure evolution during a temperature ramp up to 1,600K. The results are compared to previously published data on HBS annealing tests performed in a Knudsen cell where observed burst releases are explained as sample cracking during the experimental sequence.

Grain boundary character distribution-optimized (GBCD) Type 316 corresponding austenitic stainless steel and its cold-worked ones (GBCD+CW) are one of prospective nuclear materials to be considered for next generation energy systems. These steels were thermally-aged at 973 K for 1 and 100 h and were examined by transmission electron microscopy (TEM) to evaluate microstructural stability during high temperature exposure. TEM results revealed that microstructures of both specimens prior to ageing contained step-wise boundaries which is composed of coincidence site lattice (CSL) and random grain boundaries and also that the GBCD+CW specimens had dislocation cells and networks as well as deformation twins whereas as the GBCD one possessed few dislocations. After thermal ageing, the precipitates formed on not only random grain boundaries but also dislocations, contributing to prevent significant microstructural change occurring such as recrystallization and dislocation recovery.

We applied ab initio calculation and an object kinetic Monte Carlo modeling to the study of He-vacancy cluster nucleation under irradiation in bcc and fcc Fe, which are surrogate materials for ferritic/martensitic and austenitic steels, respectively. The ab initio calculations provided parameters for the object kinetic Monte Carlo model, such as the migration energies of point defects and the dissociation energies of He and vacancy to He-vacancy clusters. We specially focused on the simulation of high He/dpa irradiation such as He-implantation into the materials and tracked the nucleation of clusters and the fate of point defects such as SIAs, vacancies, and He atoms. We found no major difference of He-vacancy cluster nucleation between bcc and fcc Fe when we ignore the intracascade clustering even if the migration energies of point defects are significantly different between the two crystals.

The reserved cast austenitic stainless steels (CASS) for primary circuit piping in Daya Bay Nuclear Power Plant were studied. The changes of microstructure, mechanical properties and fracture behavior were investigated using SEM, EPMA, TEM and nanoindentation after accelerated aging at 400°C for up to 10000 h. Microhardness of ferrite increased rapidly in the early stage and then increased slowly later. The impact energy of materials declined with the aging time and reduced to a very low level after aging for 10000 hours. Fracture morphology displayed a mixture of cleavage in ferrite along with dimple and tearing in austenite. Two kinds of precipitations were observed in ferrite by TEM after long periods of aging. The fine Cr-enriched α′ phases precipitated homogeneously in ferrite, and a few larger G phases were observed as well. The precipitation of α′ phases was considered to be the primary mechanism of thermal aging embrittlement in CASS.

Adhesive strength of the oxidation layers on carbon steel was evaluated by means of a laser shock method, which uses a pulsed laser to generate shock wave. Oxidation for 200 hours in air created 10-micron-thick magnetite on carbon steel. Typical strength of the layer was evaluated to be about 50MPa at ambient temperature. The adhesive strength was varied from around one-tenth of yield stress to the ultimate tensile strength of the base materials. The adhesive strength of the oxide layer depended on test temperature. It is possible that the adhesive strength becomes an essential parameter for the evaluation of the protective layers.

Silicon carbide is one of the prime matrix material candidates for inert matrix fuels (IMF) which are being designed to reduce plutonium and long half-life actinide inventories through transmutation. Since complete transmutation is impractical in a single in-core run, reprocessing the inert matrix fuels becomes necessary. The current reprocessing techniques of many inert matrix materials involve dissolution of spent fuels in acidic aqueous solutions. However, SiC cannot be dissolved by that process. Thus, new reprocessing techniques are required.

This paper discusses a possible way for separating transuranic (actinide) species from a bulk silicon carbide (SiC) matrix utilizing molten carbonates. Bulk reaction-bonded SiC and SiC powder (1 μm) were corroded at high temperatures (above 850 °C) in molten carbonates (K2CO3 and Na2CO3) in an air atmosphere to form water soluble silicates. Separation of Ceria (used as a surrogate for the plutonium fissile fuel) was achieved by dissolving the silicates in boiling water and leaving behind the solid ceria (CeO2).

Radiation damage effects in ceramics, e.g., nuclear waste forms, transmutation targets, and inert matrix fuels, may have important implications for the physical and chemical stability of these materials as the cumulative radiation dose increases over time. A key aspect of scientific research in this area is the ability to understand the fundamental damage mechanisms through the combination of experimental and atomistic modelling techniques. In this paper, we review some of the lessons learned from the significant body of data now available for pyrochlore-defect fluorite based materials, followed by an illustration of the advantages of working on simple compounds with well established interatomic potentials. We conclude the paper with a description of radiation damage processes in the LaxSr1-1.5xTiO3 defect perovskites, a system that includes phase transformations, short-range order effects, and complex defect behavior.

The oxidation of uranium dioxide has been studied for more than 50 years. It was first studied for fuel fabrication purposes and then later on for safety reasons to design a dry storage facility for spent nuclear fuel that could last several hundred years. Therefore, understanding the changes occurring during the oxidation process is essential, and a sound prediction of the behavior of uranium oxides requires the accurate description of the elementary mechanisms on an atomic scale. Only the models based on elementary mechanisms should provide a reliable extrapolation of laboratory results over timeframes spanning several centuries. The oxidation mechanism of uranium oxides requires understanding the structural parameters of all the phases observed during the process. Uranium dioxide crystal structure undergoes several modifications during the low temperature oxidation that transforms UO2 into U3O8. The symmetries and the structural parameters of UO2, β-U4O9, β-U3O7 and U3O8 were determined by refining neutron diffraction patterns on pure single-phase samples. Neutron diffraction patterns, collected during the in situ oxidation of powder samples at 483 K were also analyzed performing Rietveld refinements. The lattice parameters and relative ratios of the four pure phases were measured during the progression of the isothermal oxidation. The transformation of UO2 into U3O8 involves a complex modification of the oxygen sublattice and the onset of complex superstructures for U4O9 and U3O7, associated with regular stacks of complex defects known as cuboctahedra which consist of 13 oxygen interstitial atoms. The structural modifications during the oxidation process are discussed.

The future expansion of nuclear power provides materials challenges that are not easily overcome, for example the safe immobilisation of nuclear waste is an important component in any future expansion of nuclear power. The use of ceramic-based materials, as opposed to borosilicate glasses, is now being investigated widely. The ability of ceramics to be tailored to a specific waste stream is now understood and obtainable quickly and with minimal cost. A second component that limits the expansion of fission-based technologies is the development of materials that are not only tolerant of radiation damage, but are also capable of retaining mechanical strength at high temperatures. One concern for any material however, is the effect of radiation damage, primarily from alpha-decay damage, which in many systems can transform the material from crystalline to amorphous. The effects of such radiation damage on both the structural and chemical properties can range from trivial to critical, for example volume expansion and are the primary focus of much research. As part of a long-term research programme the effects on radiation tolerance of a range of ordered and disordered materials are discused.

Even if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

Based on studies of simple oxides, this paper demonstrates that the specific energy deposition modes under irradiation induce modifications of materials over different length scales. On the other hand, we show the Landau phase transition theory, widely used to explain the structural stability of materials out of irradiation, can give a general framework to describe the behavior of these oxides under irradiation. The use of X-ray diffraction techniques coupled with the Raman spectroscopy allows defining in a quantitative way the phenomenological parameters leading to predictive results. This paper clearly shows that in two model systems, pure zirconia and spinels, no unexpected new phases are produced in these oxides irradiated at room temperature and with different fluxes. Such a phenomenological approach may be useful to study the radiation tolerance of many crystalline ceramics (e.g. the zirconium based americium ceramics).

Oxygen potentials of (Th0.7Ce0.3)O2-x were experimentally determined by means of thermogravimetric analysis as a function of non-stoichiometry at 1173 and 1273 K. Oxygen potentials of (Th0.7Ce0.3)O2-x at each temperature increased with increase of oxygen to metal (O/M) ratio (=2-x) and steep increases of the oxygen potentials when approaching O/M ratio = 2 were observed. These characteristics are typical for non-stoichiometric fluorite-type actinide dioxides. The oxygen potentials of (Th0.7Ce0.3)O2-x were similar to those of CeOO2-x when they were plotted as a function of average Ce valence.

Silicon carbide is one of the most studied materials for core components of the next generation of nuclear plants (Gen IV). In order to overcome its brittle properties, materials with nanometric grain size are considered. In spite of the growing interest for nano-structured materials, only few experiments deal with their behaviour under irradiation. To assess and predict their evolution under working conditions, it is important to characterize their microstructure and structure. To this purpose, we have studied microcrystalline and nanocrystalline samples before and after irradiation at room temperature with 4 MeV Au ions. In fact, it is well established that such irradiation conditions lead to amorphisation of the material, which can be restored after annealing at high temperature. We have performed isochronal annealings of both materials to point out the characteristics of the healing process and eventual differences related to the initial microstructure of the samples. To this purpose Grazing Incidence X-Ray Diffraction has been performed to determine the microstructure and structure parameters. We observe the amorphisation of both samples at similar doses but different annealing kinetics are observed. The amorphous nanocrystalline sample recovers its initial crystalline state at higher temperature than the microcrystalline one. This effect is clearly related to the initial microstructures of the materials. Therefore, the grain size appears as a key parameter for the structural stability and mechanical properties of this ceramic material under irradiation.

A process for the synthesis of silicon carbide (SiC) inert matrix fuels at a low temperature (1050 °C) is reported which utilized a liquid polymer precursor. As the polymer content increased, the theoretical density of the pellet at first increased and then reached a plateau. From the onset of the plateau, the packing of the one micron SiC particles in the green body was determined to be 64-68% at 600 MPa pressing pressure. As expected, mixing coarse and fine SiC particles gave a higher pellet density. The maximum density achieved was 80% of the theoretical density. Mercury porosimetry showed that the largest pore size was around 10% of the largest particle sizes present in the green body. SEM images showed that ceria, which was selected as a surrogate for PuO2 in the present study, was well distributed.

Thermal conductivities of Cs-M-O (M = Mo or U) ternary compounds, observed in the pellet-cladding gap region and in the pellet periphery in irradiated oxide fuels with high oxygen potentials, were investigated. Bulk samples of Cs2MoO4 and Cs2UO4 were prepared by hot pressing or spark plasma sintering, and their thermal diffusivities were measured by the laser flash method from room temperature to 823 K for Cs2MoO4 and to 900 K for Cs2UO4. The thermal conductivities were evaluated from the thermal diffusivity and bulk density, and the specific heat capacity values available in the literature. The thermal conductivities of Cs2MoO4 and Cs2UO4 were quite low compared with UO2 (e.g. 0.5 Wm−1K−1 at 800 K for Cs2MoO4).

The TRISO fuel has been used in some of the Generation IV nuclear reactor designs [1,2]. It consists of a fuel kernel of UOx coated with several layers of materials with different functions. Pyrolytic carbon (PyC) is one of the materials in the layers. In this study we investigate the possibility of using Glassy Polymeric Carbon (GPC) as an alternative to PyC. In this work, we are comparing the changes in physical and microstructure properties of GPC after exposure to irradiation fluence of 5 MeV Au equivalent to a 1 displacement per atom (dpa) at samples prepared at 1000, 1500 and 2000°C. The GPC material is manufactured and tested at the Center for Irradiation Materials (CIM) at Alabama A&M University. Transmission electron microscopy (TEM), Rutherford backscattering spectroscopy (RBS), X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy were used for the analysis.