Hj. Matzke

RADIATION EFFECTS IN CRYSTALLINE CERAMICS FOR THE IMMOBILIZATION OF HIGH-LEVEL NUCLEAR WASTE AND PLUTONIUM

This review provides a comprehensive evaluation of the state-of-knowledge of radiation effects in crystalline ceramics that may be used for the immobilization of high-level nuclear waste and plutonium. The current understanding of radiation damage processes, defect generation, microstructure development, theoretical methods, and experimental methods are reviewed. Fundamental scientific and technological issues that offer opportunities for research are identified. The most important issue is the need for an understanding of the radiation-induced structural changes at the atomic, microscopic, and macroscopic levels, and the effect of these changes on the release rates of radionuclides during corrosion. {copyright} {ital 1998 Materials Research Society.}

Gas release mechanisms in UO2—a critical review

Abstract The first and basic step in gas release is single gas atom diffusion. The lattice location of rare gas atoms in the UO2 lattice and the mechanism of this diffusion process are discussed. Trapping of single gas atoms is important for out-of-pile laboratory work and helps to explain much of the scatter in the reported data. It is less important in-pile due to irradiation induced resolution. The experimental techniques used to measure gas diffusion and release are critically discussed. The available data on fission gas bubble mobility are reviewed. The reliable knowledge is scarce. The effect of burn-up (change in O/M-ratio, in-growth of fission products) is treated. The relative contributions of single gas atom diffusion, fission gas bubble mobility, sweeping and inter-granular bubble growth are discussed for the cases of highly rated fast breeder fuel and moderately rated water reactor fuel. It is argued that single gas atom diffusion is contributing essentially to release in LMFBR fuel, whereas t...

Radiation damage in crystalline insulators, oxides and ceramic nuclear fuels

Abstract Studies of radiation damage in crystalline insulators usually originate from problems connected with heavy ion impact during ion bombardment, from neutron irradiation with and without fission in nuclear reactors, or from α-decay with the resulting damage due to recoil daughter atoms of the decaying nuclei of actinide compounds. The materials involved cover a broad range of compounds, e.g. from BeO to CmO2 for oxides, for which most work has been done. The damage studied ranges from production of isolated Frenkel pairs to complete amorphization of the crystalline compound (metamictization). The available knowledge is discussed. Emphasis is put on simple binary oxides and on ceramic nuclear fuel materials, i.e. oxides, carbides and nitrides of U and Pu. Recent work on irradiated glasses is also briefly discussed since these glasses are considered as promising media for safe storage of radioactive waste for long periods of time.

Effects of self-radiation damage in Cm-doped Gd2Ti2O7 and CaZrTi2O7

Abstract Specimens of Gd2Ti2O7 and CaZrTi2O7 were doped with 244Cm and the effects of self-radiation damage from alpha decay were determined as a function of cumulative dose. The macroscopic swelling of the specimens increased exponentially with dose to limiting (saturation) values of 5.1 and 6.0% for Gd2Ti2O7 and CaZrTi2O7, respectively. The radiation-induced microstructure consists primarily of individual amorphous tracks from the alpha-recoil particles which eventually overlap to produce an amorphous state at ∼ 2.0 × 10 25 alpha decays/m 3 . This radiation-induced swelling and amorphization results in increased dissolution rate and fracture toughness, but decreased hardness. Both materials recrystallize in a sharp recovery stage. The stored energy release is ∼127 J/g and the activation energy for recrystallization in CaZrTi2O7 is estimated to be 5.8 eV.

A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel: Review and recommendations

The thermal conductivity of irradiated UO2 fuel is discussed considering the effects of burnup (dissolved and precipitated solid fission products), porosity and fission-gas bubbles, deviation from stoichiometry and radiation damage based on single-effect results previously published on SIMFUEL (simulated extended burnup UO2 fuel) and on radiation damage measurements. An analytical expression including factors describing the above effects is applied to the expression for unirradiated UO2 thermal conductivity; it reflects the knowledge available today, and it is recommended for use with irradiated fuel. The expression is validated against available published data on thermal conductivity of irradiated fuel. This expression can be incorporated into fuel modelling codes to improve calculations of operating temperatures and predictions of behaviour of irradiated fuel under normal and accident conditions, including the extended burnup.

MICROSTRUCTURAL FEATURES OF SIMFUEL : SIMULATED HIGH-BURNUP UO2-BASED NUCLEAR FUEL

Abstract Simulated high-burnup nuclear fuel (SIMFUEL) replicates the chemical state and microstructure of irradiated fuel so that detailed experiments on fission-gas release, thermal conductivity and leaching can be undertaken in the laboratory. Eleven stable elements were added to simulate the compositions of 3 and 6 at% burnup UO2-based fuel. The preparation route featured high-energy grinding and spray drying to achieve homogeneous dispersion of the feed materials on a submicrometre scale. Sintering at 1650°C then provided atom-scale mixing and second-phase development characteristics of high temperature fuels. The gases and volatiles cannot be introduced during fabrication, but can be subsequently added by ion implantation. The microstructure of SIMFUEL was found to be similar to that of irradiated fuel (without bubbles). Spherical metallic Mo-Ru-Pd-Rh precipitates were found uniformly dispersed throughout the matrix. A finely-precipitated perovskite phase was observed decorating the matrix grain boundaries.

Materials research on inert matrices : a screening study

Abstract Materials research on inert matrices for U-free fuels has been extensively performed at the Institute for Transuranium Elements (ITU) for more than five years. Relevant experience, e.g. on MgO-based ceramic fuel, fabrication and irradiation of annular cercer and cermet fuel and of ThO 2 -based fuel in ITU dates back to about 30 yr ago. The criteria for selecting inert matrices for Am-transmutation, their fabrication – with and without Am – and typical results on property measurements are discussed, often in comparison with UO 2 , with emphasis on radiation damage formation and damage effects. The materials studied in most detail are spinel MgAl 2 O 4 , zircon ZrSiO 4 , ceria CeO 2− x , yttria-stabilized zirconia (Zr 1− x Y x )O 2− x /2 , monazite CePO 4 , and to a smaller degree Al 2 O 3 , MgO, SiC and Si 3 N 4 . This paper mentions and reports significant characteristics and experimental results for some of the above listed materials, as an overview of the research activities carried out at ITU. Preliminary results of first leaching experiments with Am-doped CeO 2 , MgAl 2 O 4 and ZrSiO 4 are also reported. Some recommendations deduced from this work are summarized.

On the rim effect in high burnup UO2LWR fuels

Abstract Ion implantation experiments with energetic Xe-ions were performed in order to better understand the possible mechanisms leading to the formation of an outer shell of small grains and large porosity in high burnup UO 2 (the so-called “rim effect”). It is shown that grain subdivision related to the formation of highly pressurized (probably solid) Xe-precipitates occurs at implantation doses corresponding to fission product concentrations reached at 7 to 8 at% burnup. Fracture or cleavage of UO 2 grains lead to a structure of small grains similar in size to those observed in irradiated high burnup UO 2 . The limitations of such simulation experiments are discussed and the parameters are defined for conclusive reactor irradiation experiments to finally understand the underlying mechanism.

Self-radiation damage in Gd2Ti2O7

Abstract The effects of self-radiation damage from alpha decay in Gd2Ti2O7 were investigated by studying specimens doped with 244Cm. The radiation-induced microstructure consists of individual amorphous tracks from both the alpha-recoil particles and the spontaneous fission fragments. The eventual overlap of the tracks at higher doses leads to a completely amorphous state. The self-radiation damage increases the volume, dissolution rate, and fracture toughness. Electron-beam recrystallization of the amorphous state results in the formation of fine microcrystallites on the order of 0.05 μm in size.

Transmission electron microscopy observation on irradiation-induced microstructural evolution in high burn-up UO2 disk fuel

In order to identify the conditions of the rim structure formation as a function of burn-up and temperature, and to clarify the formation mechanism of this restructuring, UO2 fuel disks were irradiated at four thermal conditions, between 400 and 1300 °C, and at four different burn-ups, between 36 and 96 MWd/kgU, without external mechanical constraint. The microstructural evolutions as a function of the irradiation parameters are observed by high resolution scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The SEM observations reveal the transition from original to sub-divided grains of rim structure and make clear that the burn-up threshold is between 55 and 82 MWd/kgU. The temperature threshold of this restructuring could be 1100±100 °C. Moreover, polyhedral sub-divided-grains with size ranging between 0.5 and 2 μm, not only rounded grains in the size range 150–350 nm, are clearly observed. These configurations are explained by assuming that the grain sub-divisions occurred homogeneously within the original polyhedral grains, while the existence of rounded grains might be due to free surface effects. TEM observations of re-structured samples show that most of sub-grain boundaries are low angle and are heavily decorated by fission gas bubbles in the range 3.5–8 nm. In the non-restructured samples, dislocations and small precipitates are present, and many of the bubbles form “strings” along dislocation lines. In specimens irradiated at high temperature, many dislocations seem to be anchored by fission product precipitates. These results suggest that the formation mechanism of the restructuring is based on polygonization, and the precipitates could have some “pinning effect” on dislocations and defect clusters.