V.V. Rondinella

The high burn-up structure in nuclear fuel

During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In todays light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

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.

Zirconate Pyrochlore as a Transmutation Target: Thermal Behaviour and Radiation Resistance against Fission Fragment Impact.

Abstract Zirconates with the pyrochlore structure (A2Zr2O7) are investigated at ITU for use as an actinide host in inert matrix fuels for transmutation. Zirconate pyrochlores with A=Nd as an inactive stand in for the trivalent actinides Am and Cm were fabricated and characterised, and their thermal transport properties measured. The low thermal conductivity indicates that zirconate pyrochlore can only be used for transmutation if it is dispersed in a cercer or cermet composite fuel. The temperature profiles of a MgO–An2Zr2O7 composite were calculated using the measured Nd-zirconate thermal conductivity for different concentrations of the included phase. The radiation stability of Nd2Zr2O7 against fission products (FP) was investigated using iodine irradiation (120 MeV). Significant alterations of the implanted regions were observed even at relatively low fluence.

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.

The thermal conductivity of Nd2Zr2O7 pyrochlore and the thermal behaviour of pyrochlore-based inert matrix fuel

Abstract High-temperature heat capacity and thermal diffusivity of neodymium zirconate (Nd 2 Zr 2 O 7 ) were measured by differential scanning calorimetry and laser flash, respectively. Measurements were made on pellets with the pyrochlore structure ( a =1.07 nm) that were fabricated from milled beads obtained by gel supported precipitation. The thermal conductivity, calculated from the measured heat capacity and thermal diffusivity, was lower than that for ZrO 2 stabilised in its cubic structure. These results show that the pyrochlore-type zirconate can be used as an inert matrix for minor actinides transmutation, but only in a composite form.

Leaching behaviour of UO2 containing α-emitting actinides

After a few hundred years in a geologic repository, α-emissions will dominate the radiation field in and around spent nuclear fuel. In the event of a failure of the spent fuel container, the dissolution of the uranium dioxide matrix in contact with groundwater could be enhanced by α-radiolysis, which could create oxidizing conditions at the fuel surface. The radioactivity (hence the radiolysis) of the irradiated fuel available nowadays is characterized by strong β- and γ-contributions, which are not representative of aged fuel in a repository. A possible way to single out the effects of α-radiolysis on the dissolution behaviour of irradiated fuel is to study unirradiated UO2 doped with a short-lived α-emitter. UO2 containing ~0.1 and ~10 wt.% of 238PuO2 was fabricated and leached. Previous results of static (batch and sequential) leaching tests on monoliths at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released were higher in the case of UO2 containing 238Pu than in the case of undoped UO2. This work presents results of leaching experiments performed under similar conditions on materials of the same composition, but crushed to give samples with two different particle size distributions to investigate the effects of different surface areas. The U release data, normalized to the geometric surface area, showed enhanced U dissolution from α-doped UO2 also in the case of crushed materials, except for the initial release after 1 h of leaching. Under these experimental conditions, the concentration of U measured in the leachates for the two doped materials did not show a clear dependence on α-activity.

α-Radiolysis and α-Radiation Damage Effects on uo2 Dissolution Under Spent Fuel Storage Conditions

α-decay will constitute almost entirely the radiation field in and around spent nuclear fuel after a few hundred years in a geological repository. Pellets of UO 2 containing ˜0.1 and ˜10 wt. % 238 Pu were fabricated using a sol-gel method and characterized, comparing their properties to those of undoped UO 2 . The α-radiation fields of different types of commercial LWR spent fuel are of the same order of magnitude as the fuel with the lower Pu-concentration used in this work. The results of static batch leaching tests at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released during leaching were higher in the case of UO 2 containing 238 pu than for undoped UO 2 . Relatively large amounts of Pu were released after the longest leaching times. Lattice parameter measurements using XRD and hardness measurements by Vickers indentation showed a relatively rapid build-up of α-decay damage in the material stored at ambient temperature with the higher concentration of dopant, while for the material with ˜0.1 wt. % Pu no clear variations were detected during the same time intervals.

Damage produced in magnesium aluminate spinel by high energy heavy ions including fission products of fission energy: microstructure modifications

The radiation stability of spinel MgAl 2 O 4 against the impact of beams of a fission product at fission energy (70 MeV iodine) and at different fluences was investigated using Transmission Electron Microscopy (TEM). Specimens prethinned before irradiation were analysed by TEM and irradiated bulk specimens were investigated using cross-sectional TEM. Tracks were observed in pre-thinned specimens. Partial amorphisation was observed for the irradiation at the highest fluence. Recrystalization of the amorphous region, induced by the microscope electron beam was observed. A threshold value of 6 keV nm -1 was determined for the amorphization of spinel under the above mentioned irradiation conditions. Moreover, profile measurements of the irradiated areas confirmed the large swelling values for this material when irradiated with fission products of fission energy. A thermal spike model was used to calculate the damage threshold for spinel using experimentally measured heavy-ion track radii, including results for other ions of up to very high energies. These accelerator-based fission product irradiations revealed an unexpected poor radiation stability, in contrast to the known good behaviour of the spinel against neutron or alpha particle damage.

Development of Fast Reactor Metal Fuels Containing Minor Actinides

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phénix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU.

Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel.

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.