T. Wiss

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.

Swift heavy ion and fission damage effects in UO2

Abstract Irradiations of uranium dioxide, UO2 with different swift heavy ions were performed using wide ranges of energies and fluences from 72 MeV to 2.7 GeV in the range 5×109–1017 ions/cm2. The ions were Zn, Mo, Cd, Sn, Xe, I, Pb, Au and U. The threshold energy loss for formation of visible tracks in UO2 could be determined to be in the range 22–29 keV/nm. Fission products of fission energy are below this threshold but nevertheless form thermal spikes in UO2. Observable tracks are only found at the surface. By using 127I-beams of 72 MeV energy the consequences of fission product impact, i.e., lattice parameter increase, fission gas bubble formation, resolution of fission gas from bubbles and fission-enhanced diffusion were all observed and measured. The swelling of UO2 was confirmed to be small and the technologically important process of polygonization (grain subdivision, also called rim-effect in operating UO2-fuel) could be simulated. The results obtained are important in understanding the operating behavior of UO2, today’s fuel in nuclear electricity-producing power stations.

Radiation damage in UO2 by swift heavy ions

Abstract Specimens of sintered UO2, a high melting point ceramic with the fluorite structure, were irradiated with heavy ions (129Xe, 238U) with different fluences (5 × 1010 to 7 × 1013 ions/cm2) and energies (173 MeV for Xe ions to 2.713 GeV for U ions). The influence of the electronic energy loss on the mechanisms of damage formation was studied in the range of 29 to 60 keV/nm. Transmission Electron Microscopy (TEM) was performed to identify and characterize the damage induced by these ions. Tracks produced by U ions of 2713 and 1300 MeV and by Xe ions of 173 MeV were observed. The radii of the observed tracks were calculated using a thermal-spike model, taking into account the thermodynamic parameters of the material and the energy and velocity of the incoming ions. The TRIM code was used to determine the displacement profile and the energy distribution along the ion paths. Good agreement with the experimental results was found. The dependence of damage formation on the ion dose was also studied. For instance, defect clusters and loops were produced in UO2 irradiated with 129Xe of 173 MeV ( d E d x ∼ 29 keV/nm ) between 7 × 1010 and 7 × 1013 ions/cm2.

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.

Predicting material release during a nuclear reactor accident

In the aftermath of a nuclear reactor accident, understanding the release of fission products from the fuel is key.