Advances in Science and Technology Vol. 73

Abstract: Ceramic matrix composites (CMC) are very attractive materials for structural applications at high temperatures. Not only must CMC be damage tolerant, but they must also allow thermal management. For this purpose heat transfers must be controlled even in the presence of damage. Damage consists in multiple cracks that form in the matrix and ultimately in the fibers, when the stresses exceed the proportional limit. Therefore the thermal conductivity dependence on applied load is a factor of primary importance for the design of CMC components. This original approach combines a model of matrix cracking with a model of heat transfer through an elementary cracked volume element containing matrix crack and an interfacial crack. It was applied to 1D composites subject to tensile ant thermal loading parallel to fiber direction in a previous paper. The present paper compares predictions to experimental results.

Abstract: The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S), the Ti-additional type of PNC1520 (1520Ti) and the Zr-additional type of PNC1520 (1520Zr) by using a supercritical water autoclave. In view of general corrosion, 1520Zr may have larger possibility than 1520S and 1520Ti to adopt a supercritical water reactor core fuel cladding.

Abstract: Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235U), low-density UAlx research reactor fuel with high-density, low enriched (<20% 235U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U3Si2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U3Si2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

Abstract: The experimental Electron Probe Micro Analysis (EPMA) characterizations on the MOX fuels evidence a heterogeneous microstructure, containing several phases. This heterogeneity must be accounted for in the numerical simulation. The first phase of this work, presented here, concerns exclusively the numerical representation of the MOX microstructure in three dimensions. Three identified steps were realized. The first one consisted in the acquisition and the treatment of two-dimensional experimental pictures thanks to a soft-ware already developed [1]. From the made treatments, the following bi-dimensional data were acquired: the surface fraction of every phase, the various diameters of inclusions within a phase as well as their surfaces fractions. However, within the framework of our study, we wished to represent our heterogeneous microstructure in three dimensions. Except, the data, supplied by this soft-ware, were bi-dimensional. Therefore, the second step of our works deal with the stereological domain. The model of Saltykov [2] was used to go back up the two-dimensional statistical information in three-dimensional. Finally, the last step of our works was to develop a tool able to build a meshed periodic numerical representation of the MOX microstructure. This innovative tool, based on a Random Sequential Absorption technique, represents MOX fuels already irradiated in reactor or any heterogeneous fuels envisaged in the future as well. For example it models two or three phases MOX fuel or any multi-phases fuels as well. Moreover, the sizes of the inclusions can vary within each phase. At the moment, the tool models spherical inclusions but nothing prevents from evolving towards more complex morphologies.

Abstract: Fuels for future fast reactors will not only produce energy, but they must also actively contribute to the minimisation of long lived wastes produced by these, and other reactor systems. The fuels must incorporate minor actinides (MA = Np, Am, Cm) for neutron transmutation into short lived isotopes. Within Europe oxide fuels are favoured. Transmutation can be considered in homogeneous or heterogeneous reactor recycle modes (i.e. in fuels or targets, respectively). Fabrication of such fuels can be made by advanced liquid processing methods, enabling property determination and screening irradiation experiments. This paper will describe these fabrication processes, and discuss properties and fuel irradiation experiments made to date. Both fertile and inert matrix fuel types are considered.

Abstract: This work is devoted to the study of the geometrical stability in time of UyAm1-yO2-x pellet due to self-irradiation at room temperature. Dense and tailored porosity U1-yAmyO2-x (y=0.10; 0.15) compounds were fabricated by a powder metallurgy process. Up to 3.4.1017 α particles/g, an increase in the diameter of 0.9% was observed for the dense compounds while for the tailored porosity material a diameter increase of 0.4% was observed. Swelling laws have been established from the experimental data.

Abstract: Experiments and atomic-scale computer simulations have shown that nano-scale voids and copper precipitates can be strong obstacles to the glide of dislocations in neutron-irradiated iron. Simulations have shown that voids are strong obstacles and that an edge dislocation climbs by absorbing vacancies at it breaks away from voids. The obstacle strength of copper precipitates is enhanced by a dislocation-induced structural transformation if they are large enough and the temperature is low enough. Most simulations have the centre of a spherical void or precipitate on the slip plane of an edge dislocation. The present work investigates how the strength of 2 and 4 nm voids and precipitates varies with the distance of their centre from the slip plane at temperatures across the range 0 to 450 K. The strength of voids is highest when their centre coincides with the slip plane, but this is not the case for small precipitates, which do not transform from the bcc structure. The strength of both type of obstacle, and the extent of climb at voids and transformation of large precipitates are not symmetric with respect to the position of their centre from the slip plane. The results are discussed in terms of the atomic mechanisms involved.

Abstract: The objective of this work is to evaluate the microstructure of the neutron-irradiated reduced activation ferritic/martensitic (RAFM) steel EUROFER 97. For this purpose irradiation induced defects like defect clusters, dislocation loops, voids/bubbles and precipitates are identified by transmission electron microscopy (TEM) and quantified in size and volume density. Emphasis is put on analyzing the influence of the irradiation dose and neutron fluxe on the evolution of size and density of the defects at irradiation temperatures between 300 and 335 °C. A first sample irradiated to a dose of 31.8 dpa was analyzed. The irradiation was carried out in the BOR 60 fast reactor of JSC “SSC RIAR” in Dimitrovgrad, within the framework of the ARBOR-1 irradiation program. To study the dose dependence in a next step the results will be compared to quantitative data on samples irradiated to a dose of 15 dpa. The obtained quantitative data will be used for correlation of the changes in the microstructure to the changes in the mechanical properties and will serve as an input for models describing this correlation.

Abstract: The influence of helium on the mechanical properties of reduced-activation ferritic/martensitic Cr-steels under fusion-relevant irradiation conditions is still a concern. While the fact that He can influence the mechanical properties is well established [1,2], the underlying mechanisms are not fully understood [1,2]. In this work the effect of He and displacements per atom (dpa) on the irradiation-induced hardening of Eurofer97 at 300°C was studied. Self-ion irradiation was applied to simulate the neutron-irradiation-induced damage. Helium was implanted prior to (pre-implantation), simultaneously (dual-beam irradiation) or following the (post-implantation) self-ion irradiation to investigate the He effect. Nanoindentation was used in order to characterize the damage layer. Under the present conditions (300°C, 1 dpa, 10 appmHe) the observed hardening increased in the following order: single-beam Fe-ion irradiation/pre-implantation < simultaneous implantation < post-implantation. We conclude, that there is a significant interaction between damage and He. Additionally, Eurofer97 and ODS-Eurofer were irradiated with Fe ions up to 1 and 10 dpa to study the effect of the oxide particles on the irradiation-induced hardening. We have found a higher irradiation-induced hardening at 1 dpa for ODS-Eurofer but a steeper hardness increase per dpa up to 10 dpa for Eurofer97.

Abstract: Hot isostatic pressing (HIP) is a technology with wide applicability in consolidating calcined intermediate-level and high-level nuclear waste, especially with wastes that are not able to be readily processed by vitrification at reasonable waste loadings. The essential process steps during the HIP cycle will be outlined. We have demonstrated the effective consolidation via HIP technology of a wide variety of tailored glass-ceramic and ceramic waste forms, notably simulated ICPP waste calcines, I sorbed upon zeolite beads, Pu-bearing wastes, inactive Cs/Sr/Rb/Ba mixtures, simulated waste pyroprocessing salts from spent nuclear fuel recycling, Tc, U-rich isotope production waste, and simulated K-basin (Hanford, WA, USA) and Magnox sludges (UK). Can-ceramic interactions have been carefully studied. The principal advantages of the HIP technology include: negligible offgas during the high temperature consolidation step, relatively small footprint, and high waste loadings. As a batch process, the wasteform chemistry can be readily adjusted on a given process line, to deliver wastes into different end states (e.g. direct HIP versus chemically tailored). This flexibility allows the treatment of multiple waste streams on the one process line.