The molecular dynamics (MD) calculation was performed for minor actinide (MA: Np and Am)-containing mixed oxide (MOX) fuels, U0.7-xPu0.3MAxO2, in the temperature range from 300 to around 2500 K to evaluate the thermal expansion, heat capacity, and thermal conductivity. The MD results showed that the calculated heat capacity and thermal conductivity were similar in all the composition ranges, indicating that MA scarcely affected the thermal properties of the MOX fuel in the perfect crystal system.

Mechanical properties of titanium, zirconium, and hafnium hydrides were evaluated by indentation tests in the present study. Single-phase bulk specimens of titanium, zirconium, and hafnium hydrides with a fluorite type structure were produced using a Sieverts' apparatus. The indentation hardness values of titanium and hafnium hydrides were lower than those of the pure metals and drastically decreased with increasing hydrogen content. The hardness of zirconium hydride was higher than that of pure zirconium and slightly depended on hydrogen content. The hardness of the hydrides decreased with increasing maximum indentation load, which has been referred to as indentation size effect. The dependences of hardness on the indentation depth for the hydrides were found to obey the Nix–Gao model. The indentation size effects for the hydrides were significantly smaller than those for the pure metals.

SPS technique was applied for fabrication of UN pellets. The dense sample was obtained with relatively low sintering temperature without any milling process and sintering additive, indicating that the SPS technique is suitable for fabrication of nitride fuel. The thermal conductivity and Young's modulus in axial direction were higher than those of reported values. EBSP analysis and electronic calculation supposed that weak grain orientation in [100] direction of the sample lead to these higher values.

The development of a new generation of nuclear reactors (Gen-IV), with improved thermodynamic yield and a reduction of waste production, makes necessary to consider materials able to withstand high operating temperatures. Transition metal carbides, like ZrC, are then under consideration. Despite their good thermal and neutron properties, they have unfortunately a brittle mechanical behaviour. This is the reason why it is important to investigate the properties of these systems with sub-micrometric grains and as a function of their composition. Therefore, samples having micrometric and nanometric grain sizes (and different oxygen content) were irradiated by low energy ions at room temperature to simulate their behaviour in a neutron flux. The irradiation effects in these materials were studied by grazing X-ray diffraction and transmission electron microscopy.

Oxidation behaviors of Zr-Nb alloys were investigated under the condition of high temperatures and dry air.

Experimental charged-particle irradiations and multi-scale computer simulations have been used to investigate the primary damage state and evolution of damage in silicon carbide as functions of temperature and charged-particle mass and energy. Atomistic simulations of energetic C and Si collision cascades, similar to those created by reactor neutrons, indicate that single interstitials, vacancies, antisite defects, and small defect clusters are produced. The point defects are dominated by close Frenkel pairs, and atomistic simulations indicate that the activation energies for recombination of most close pairs range from 0.24 to 0.38 eV, which suggest significant reduction in defect survivability at room temperature. Atomistic simulations have also determined that the activation energies for long-range diffusion of C and Si interstitials are 0.7 and 1.5 eV, respectively. Using these activation energies and ab initio results as input parameters, a kinetic Monte Carlo (MC) simulation model has been developed to study isochronal annealing of defects in SiC between cascade events. The defects are produced by a 10 keV Si cascades in a molecular dynamics (MD) simulation cell, and these defects are then accurately transferred to defect lattice sites in the Monte Carlo model to investigate defect recovery. By transferring defects states back and forth between the MD and MC environments, damage accumulation can be investigated as a function of temperature. Charged particle irradiations are often used to simulate radiation damage from neutrons and radioactive decay; however, at extreme charged-particle fluxes used in irradiation studies to simulate radiation damage in nuclear materials, the ratio of ionization rate to displacement rate can have a significant impact on observed temperature-dependent processes, which can affect both interpretation and model development. At high charged-particle fluxes, the defect recovery rates in SiC increase nearly linearly with the ratio of ionization rate to displacement rate. A fundamental understanding of these ionization effects is needed if charged particle irradiation results are to be used to develop predictive models of damage evolution in nuclear materials, such as SiC, as functions of time, temperature and dose rates.

We made various nitride fuels containing simulated FP elements and evaluated the effect of these FP elements on the properties of the nitrides. For Uranium Neodymium nitride solid solution the lattice parameter increased with Nd content, thermal expansion coefficient did not change and thermal conductivity decreased with Nd content. The thermal expansion for Pd containing UN, where Pd precipitated as UPd3 in the grain boundaries of UN, was nearly the same as that of UN and the thermal conductivity for Pd containing in the UN matrix decreased with Pd content. For Mo containing UN Mo precipitated as Mo metal isotropically. Both the thermal expansion and thermal conductivity did not vary with Mo content This might result from the low Mo contents at these simulated burnups.

Renewed interest in nuclear power in the United States has prompted investigations into new reactor designs, resulting in a need to gain a greater understanding of the properties of the materials which are proposed for use in next generation nuclear reactors. This presentation will focus on preliminary results of large-scale empirical potential atomistic studies into the generation of point defect clusters in 3C SiC by particle irradiation and the evolution from point defect clusters to ‘voids’ on the atomic scale. Our working definition of ‘void’ will be explained in the context of small length-scale simulations. The determination of interstitial and vacancy diffusivities for the empirical potential employed and its impact on defect coalescence will be discussed. The characterization of initial damage states for given irradiation conditions will be presented and compared to previous work on ceramics and ceramic-composites.

Oxide ceramics with fluorite structure attracts considerable attention as a host of inert matrix fuels and a transmutation target for minor actinides and long life fission products. We will report the characteristic features of defect clusters in oxide ceramics with fluorite structure, such as yttria stabilized zirconia (YSZ: 13 mol% Y2O3-ZrO2), ceria (CeO2), which were investigated in situ by transmission electron microscopy under electron irradiation. Results on calcium fluorite (CaF2) will be also shown for a comparison [1]. Electron irradiation with 100 to 1000 keV was found to induce large defect clusters in YSZ, which accompany strong black/black lobes contrast and multiply dislocation networks when they grow to a critical diameter of about 1.0 ìm. The defect clusters were considered to be charged oxygen platelets formed by the selective displacement damage of oxygen sublattice. A large difference in mass of constituent ions (O and Zr/Y ions) attributes the selective displacement damage under electron irradiation [2]. Similar defect clusters with strong diffraction contrast were formed in CaF2 under 200 keV electron irradiation, and they also multiplied dislocations at a critical diameter [1]. In the case of CeO2, a higher density of defect clusters was formed compared with YSZ and CaF2. Defect clusters formed under electron irradiation less than 1250 keV were found to be faulted-prismatic-interstitial loops lying on {111} planes with Burgers vector parallel to <111> directions [3], whereas electron irradiation with 1500, 2000 and 3000 keV induces perfect dislocation loops of 1/2<110>{110} in nature. Formation of radiation-induced defects is found to depend on electron energy for crystals with fluorite structure investigated in this study. Results will be discussed through selective displacement damage and charge accumulation. This work was financially supported by the Budget for Nuclear Research of MEXT, based on the screening and counseling by the Atomic Energy Commission, and by the Grant-in-Aid for Scientific Research (B) from JSPS. [1] M. Watanabe, T. Noma, K. Yasuda, K. Yasunaga, S. Matsumura and C. Kinoshita, Proc. of 16th Electron Mmicroscopy Congress, (2006) 1854. [2] K. Yasuda, C. Kinoshita, S. Matsumura, A.I. Ryazanov, J. Nucl. Mater., 319 (2003) 74. [3] K. Yasunaga, K. Yasuda, S. Matsumura and T. Sonoda, Nucl. Instr. and Meth. B, 250 (2006) 114.

Young's moduli of hydrogenated Zr-Nb alloys were evaluated by using a multiple elastometer in the temperature range from room temperature to 773 K, based on the cantilever characteristic vibration technique. The decreasing rate of Young's modulus against the temperature was independent on the additive amount of Nb. The Young's modulus for the hydrogen solid solution of Zr-Nb alloys decreased and that those for the alloys with partially precipitated hydride slightly changed with increasing hydrogen content in both of Zr-1.0Nb and Zr-2.5Nb. The decreasing rates with hydrogen content of hydrogen solid solution of Zr-Nb alloys were almost same as that of Zr hydrogen solid solution. This result was considered because all specimens largely consisted of α-zirconium in which the solute hydrogen was found to decrease the Young's modulus.

We investigated the synthesis of bismuth oxy-iodide and iodate compounds, in an effort to develop materials for iodine recovery from caustic waste streams and/or final waste disposal if repository conditions included ambient conditions similar to those under which the iodine was initially captured. The results presented involve the in-situ crystallization of layered bismuth oxide compounds with aqueous dissolved iodine (which resides as both iodide and iodate in solution). Although single-phase bismuth oxy-iodide materials have already been described in the context of capturing radioiodine, our unique contribution is the discovery that there is a mixture of Bi-O-I compositions, not described in the prior work, which optimize both the uptake and the degree of insolubility (and leachability) of iodine. The optimized combination produces a durable material that is suitable as a waste form for repository conditions such as are predicted at the Yucca Mountain repository (YMP) or in a similar type of repository that could be developed in coordination with iodine production via Global Nuclear Energy Program (GNEP) production cycles.

Radiation tolerant ceramics are needed to advance the utilization of nuclear energy to meet rising global energy demand. They have potential to meet the demanding radiation environments in applications as nuclear fuel and waste forms. Their discovery and development are hampered by a lack of fundamental understanding of the physics underlying radiation tolerance of ceramics. Several theories have been advanced in the literature based on structure, radius ratio of cations and ionicity or covalency of bonds. Most of these theories focus on defect production and accumulation processes only, but neglect in-cascade and thermal annihilation of defects. We have performed large scale molecular dynamics simulations of 30 keV U and Zr recoils in zircon (ZrSiO4), zirconia (ZrO2) and yttria-stabilized zirconia (YSZ) to understand the atomic-level mechanisms that contribute to radiation tolerance, particularly fast annihilation processes. Zircon is amorphized by irradiation, while zirconia and YSZ do not undergo radiation-induced amorphization. Our results reveal that dynamic defect annihilation is very effective at controlling defect accumulation in radiation tolerant materials. Based on our results, we will discuss a strategy for improving the radiation tolerance of ceramics.

Carbide ceramics as SiC, TiC or ZrC are potential candidates for high temperature applications such as fourth generation nuclear plants because of their refractory or low activation under neutron irradiation properties. Nevertheless, the typical drawbacks of hard ceramics (brittleness) could limit their use in these applications. In order to overcome these problems, one possibility is to decrease the grain size down to the nanometric scale. Enhancement of the mechanical properties is actually expected in such nanostructured ceramics (ductility) and moreover, these nanomaterials could also take advantage of their strong grain boundaries density to withstand severe irradiation conditions. If one wants to quantify the expected enhancement of the properties, the first challenge that has to be faced is the elaboration of the nanostructured ceramics samples. That means being able to synthesize the pre-ceramics nanopowders in weighable amounts, and then finding an efficient way to sinter them aiming at the maximum densification together with avoiding grain growth.

In this contribution, we present SiC, TiC and ZrC nanopowders synthesis by laser pyrolysis and inductively coupled plasma, together with their densification by different techniques (Hot Isostatic Pressing, Spark Plasma Sintering, High Pressure Flash Sintering). We also report the latest findings obtained on the behavior of SiC nanostructured ceramics under low energy ion irradiation.

Raw micrometric SiC and ZrC powders were used as precursors in the inductively coupled plasma experiment. The production was as high as 1 kg.h-1, with nanograins ranging from 10 to 100 nm in size depending on the synthesis conditions. For the laser pyrolysis method, gaseous precursors (SiH4, C2H2) were used for SiC while liquid alkoxides precursors were used for TiC and ZrC respectively. For SiC, the production rate can reach 100 g.h-1 (laboratory scale) with grain sizes ranging from 10 to 50 nm with narrow size distribution. For TiC and ZrC nanopowders, the production rate is lower than for SiC because of the use of liquid precursors that leads to a worse yield. In this latter case, the carbide phase is obtained after carburization of the laser pyrolyzed TiO2 (or ZrO2) / free carbon nanocomposites. The final carbide nanograins size is in the 50 – 80 nm range. After sintering, the obtained pellets show different characteristics depending on the starting powder and the sintering technique. With the right sintering conditions, the densification reaches 95 % without any sintering additives, with no (or limited) grain growth and no modification of the crystalline structure. Concerning the properties of the obtained nanostructured ceramics, the SiC pellets, together with the as-synthesized nanopowders, were submitted to low energy ion irradiation in order to compare their behavior to conventional SiC materials.

CeO2 thin films were irradiated with 200MeV Xe ions. Effects of the irradiation were studied by using Extended X-ray Absorption Fine Structure (EXAFS) measurement at SPring8 synchrotron radiation facility. EXAFS spectra for the irradiated thin films near the Ce K-edge show that the coordination number for oxygen atoms around Ce atom decreases and that the Ce-O Debye-Waller factor increases by the irradiation. The atomic distance between oxygen atom and Ce atom does not vary within the accuracy of EXAFS measurement. The effect of high density electronic excitation on the structure of CeO2 is discussed.

The main problem we have to face in order to improve the in-pile behaviour of the UMo/Al dispersed fuel, is the instability of the interaction zone, that is formed between the UMo particles and their surrounding Al matrix during irradiation. As the latest irradiations (e.g. IRIS 3 or RERTR 6) tend to prove, addition of Si into the Al matrix seems to be beneficial for the interaction rate decrease. To further understand the role of silicon, a wide out-of-pile interdiffusion study has been launched in CEA-France in 2006 and is still under progress. It has given promising preliminary results by showing a real effect of Si addition upon the interaction zone. The first results related to the physico-chemical aspects of the modified UMo/Al interaction as well as the kinetics aspects of the interaction layer (IL) formation are described in this paper. On a wide range of annealed couples made of UMo7 or UMo10 and Al-Si matrix (Si content ranging from 0.11 wt % to 12 wt %), the IL thicknesses have been systematically measured. Microstructures have been fully characterized by means of optical and scanning electron microscopy and the elementary compositions have been evaluated through EDX analyses. Finally, XRD characterizations and micro-hardness tests have also been performed on a few samples. On the basis of these different results, we discuss first clues about the role played by silicon on slowing down the UMo/Al interaction process.

The proposed designs of GenIV reactors call for high operating temperatures and long exposure times, aimed at increasing thermal efficiency, enabling hydrogen production, decreasing waste and increasing safety, among other goals. These requirements will require imply the materials will be subjected to higher radiation damage doses than seen in light water reactors, requiring an even higher degree of alloys stability. In addition the higher temperatures and exposure times in corrosive environments will require much higher corrosion resistance than current materials. One of the main challenges of designing these materials is that our ability to predict material response in the face of the synergistic effects of temperature, radiation damage and a corrosive environment is limited. However, efforts are underway to understand mechanistically the degradation processes so that they can be extrapolated to conditions beyond the experimental database.

In this context, it is interesting to consider the effort made by the nuclear power industry to qualify nuclear fuel for operation at higher burnup in existing light water reactors. In the last two decades, the average discharge burnup of nuclear fuel in light water reactors has increased almost two fold, thereby increasing the reactor exposure time and the amount of radiation damage withstood by the cladding. Experience has shown that the material degradation rates can increase at high burnup, sometimes through the synergistic operation of various processes. Two examples will be given to illustrate the complexity of the problem:

(i) The rate of irradiation growth of zirconium alloys increases significantly at high burnup. The mechanism has been shown to depend on the amorphization of intermetallic ZrCrFe precipitates with consequent release of Fe into the Zr matrix. This has been shown to help nucleate component dislocation loops which preferentially absorb vacancies, thereby increasing the net rate of growth. This causes the interstitial and vacancy fluxes both to contribute towards irradiation growth, thus increasing the rate.

(ii) One of the most significant obstacles to approval of cladding operation at high burnup is difficulty in the evaluation of the behavior of the fuel during a reactivity initiated accident (RIA). The degradation of the mechanical properties of the zirconium cladding with increasing corrosion and consequent hydrogen ingress, the radiation damage and the microstructure evolution in the fuel all have to be taken into account in evaluating the likelihood of severe fuel failure at high burnup. This problem is still being addressed in various research programs.

Such mechanisms and the efforts to qualify nuclear fuel for higher burnup in light water reactors will be reviewed as a means of illustrating the challenges (known and unknown) faced by GenIV reactor materials.

Modeling cascade and fission damage evolution of actinide materials of all kinds is essential for understanding their aging characteristics. As an example of how exotic some of the damage evolution behavior can be, plutonium-gallium (Pu-Ga) alloys in the δ-phase (fcc lattices) are explored. Aging emanates from the wide variety of spontaneous decay and fission products that, in the case of the Pu, are such species as helium (He) and uranium, among others, as well as interstitials, and vacancies. To aid in our understanding, the modified embedded atom method (MEAM) formalism is applied to the Pu-Ga-He system. The behavior of defects in the fcc (δ) phase of Pu-based materials is strongly influenced by the metastability of this phase. The influence of this metastability on minimum displacement threshold energy, point defect characteristics and He bubbles is delineated. The roles of short-range ordering and transformations of voids into stacking fault tetrahedra in the aging process are also examined.

Thermodynamic modeling has been performed for O-Am binary system and O-Pu-Am ternary system on the basis of thermodynamic data available in the literature. The interaction parameters of the excess Gibbs energies of the fluorite structure FCCC1 phase in the O-Am and the O-Pu-Am systems were evaluated with respect to the deviation from the ideal solution in oxygen potential, respectively. The thermodynamic modeling used in this study is based on the CALculation of PHAse Diagram (CALPHAD) technique.

To improve understanding of radiation damage and recovery process, especially under condition of high energy and high fluence irradiation, is currently studied at the new cross-over project (NXO). Most severe irradiation is realized by fission fragments in nuclear fuel. A guiding experiment is taken from experience in power generating Light Water Reactor (LWR) fuel. At high burnup around 7%FIMA, fuel ceramics do have restructuring of the grain sub-division where diameter of grains change down to 50 to 200 nm. The NXO investigates the process mechanism crossing over research activities of universities, national and private laboratories. Simulation studies are being performed to find principal and triggering processes, using accelerator irradiation and computational calculations. Accelerator irradiation partly succeeded to reproduce the process outside of the fission reactor using simulation material of CeO2. High resolution TEM observations, comparing microstructure changes of high burnup reactor fuel and the simulated material, shows that principal process is polygonization and indicates that Oxygen defects and planar structures, of different scales, have key role on the kinetics. Basic research works of computing science, including first principle calculations, molecular dynamics, Monte-Carlo, and meso-scale cellular automata modeling are underway.

Framework of computational study is based on understanding of the repeat of passing of fission-fragment tracks, which provide overheating and quenching cycles in time domain. It throws local atoms into higher energy quasi-stable placement and configurations. Extensive experimental observations, previously obtained and presently gained by accelerator irradiations, enable the analyses to focus into key configurations in search of the target process. Stability of planar precipitation of Xenon atomson [111] plane of the fluorite structure were demonstrated by molecular dynamics calculation. Complex kinetic processes, including stability and collective behavior, of interstitial Oxygen atoms are investigated by first principle calculations, molecular dynamics and Monte-Carlo studies.

Statistical irradiation-dynamics, for studying repeated energy-deposit cycles, may enable to draw kinetic phase diagrams. Discussions were made to have statistical counting of possible repeated traces, in relevant narrowed “investigation range” defined in time domain, configuration space and local energy.