Fission Fragment Damage Research Articles

The effect of ion irradiation on the dissolution of UO2 and UO2-based simulant fuel

The aim of this work was to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO2 and UO2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO2 samples, observed by scanning electron microscopy, was to induce a smoothening of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards ∼10−9 mol/l after a few hours. Apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or ‘effective solubility’ was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO2 samples after the dissolution experiment.

The effect of fission-energy Xe ion irradiation on dissolution of UO2 thin films

The aim of this work was to study the effect of fission fragment damage on the dissolution of UO2 thin films in water. For this purpose, thin films of UO2 on LSAT (Al10La3O51Sr14Ta7) substrates were produced and irradiated by 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate the fission damage and induce chemical mixing that occur within nuclear fuels. The dissolution experiment was conducted under a nitrogen atmosphere (200–900 O2 ppm in N2) to study the effect of the induced irradiation damage and mixing on the dissolution of the UO2 matrix. The irradiated samples showed a decrease in the amount of dissolved uranium, as compared to the corresponding unirradiated samples. This was ascribed to the irradiation-induced chemical mixing of the UO2 films with the substrate elements, which resulted in stabilisation of the UO2 matrix and increased its aqueous durability. Secondary phases were also observed on the surface of the UO2 films after the dissolution experiment.

Heavy ion linear accelerator for radiation damage studies of materials.

A new eXtreme MATerial (XMAT) research facility is being proposed at Argonne National Laboratory to enable rapid in situ mesoscale bulk analysis of ion radiation damage in advanced materials and nuclear fuels. This facility combines a new heavy-ion accelerator with the existing high-energy X-ray analysis capability of the Argonne Advanced Photon Source. The heavy-ion accelerator and target complex will enable experimenters to emulate the environment of a nuclear reactor making possible the study of fission fragment damage in materials. Material scientists will be able to use the measured material parameters to validate computer simulation codes and extrapolate the response of the material in a nuclear reactor environment. Utilizing a new heavy-ion accelerator will provide the appropriate energies and intensities to study these effects with beam intensities which allow experiments to run over hours or days instead of years. The XMAT facility will use a CW heavy-ion accelerator capable of providing beams of any stable isotope with adjustable energy up to 1.2 MeV/u for 238U50+ and 1.7 MeV for protons. This energy is crucial to the design since it well mimics fission fragments that provide the major portion of the damage in nuclear fuels. The energy also allows damage to be created far from the surface of the material allowing bulk radiation damage effects to be investigated. The XMAT ion linac includes an electron cyclotron resonance ion source, a normal-conducting radio-frequency quadrupole and four normal-conducting multi-gap quarter-wave resonators operating at 60.625 MHz. This paper presents the 3D multi-physics design and analysis of the accelerating structures and beam dynamics studies of the linac.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Molecular Dynamics Simulation of Fission Fragment Damage in Nuclear Fuel and Surrogate Material

ABSTRACTWe have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO2) for energy deposition per unit length of 3.9 keV/nm. We did not observe amorphization. The damage mainly consisted of isolated point defects. Only about 1% of the displacements occur on the uranium sublattice. Oxygen Frenkel pairs are an order of magnitude more numerous than uranium Frenkel pairs in the primary damage state. In contrast, previous results show that the ratio of Frenkel pairs on the two sublattices is close to the stoichiometric ratio in ceria. These differences in the primary damage state may lead to differences in radiation response of UO2 and CeO2.

The effect of fission-energy Xe ion irradiation on the structural integrity and dissolution of the CeO2 matrix

This work considers the effect of fission fragment damage on the structural integrity and dissolution of the CeO2 matrix in water, as a simulant for the UO2 matrix of spent nuclear fuel. For this purpose, thin films of CeO2 on Si substrates were produced and irradiated by 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate fission damage that occurs within nuclear fuels along with bulk CeO2 samples. The irradiated and unirradiated samples were characterised and a static batch dissolution experiment was conducted to study the effect of the induced irradiation damage on dissolution of the CeO2 matrix. Complex restructuring took place in the irradiated films and the irradiated samples showed an increase in the amount of dissolved cerium, as compared to the corresponding unirradiated samples. Secondary phases were also observed on the surface of the irradiated CeO2 films after the dissolution experiment.

Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained.

Heat transfer simulations of the UO2 particle–graphite system in TREAT fuel

In this study, a heat transfer simulation of a UO2 particle–graphite system in highly enriched nuclear fuel at the Transient Reactor Test Facility (TREAT) was performed using the finite element method. Different factors that can impact fuel performance were modeled and implemented in the simulated micro-scale UO2 particle–graphite system. The fission fragments caused an irradiation-induced degradation of the thermal conductivity of the graphite, which added major heat resistance to the irradiated system. The effect of graphite quality and irradiation on the UO2 particles has also been evaluated, but neither has an impact as pronounced as the fission fragment damage to the graphite. By combining these factors, the dynamic temperature profiles were obtained, and the limitations on particle size in the irradiated and unirradiated UO2 particle–graphite systems have been determined.

Effects of swift heavy ion irradiation on the structure of Er 2O 3-doped CeO 2

In order to simulate the effects of burnable poison doping on the fission fragment damage of UO 2 nuclear fuels, Er 2O 3-doped CeO 2 pellets were irradiated with 200 MeV Xe 14+ ions. The irradiation effect was measured by means of X-ray diffraction (XRD). The expansion of lattice and the disordering of atomic arrangement due to the irradiation become more remarkable with increasing the concentration of the Er 2O 3 dopant.

Structural modifications in pyrochlores caused by ions in the electronic stopping regime

Damage caused by swift heavy ions with three different electronic stopping powers (120 MeV Au 9+ (22 keV/nm), 90 MeV I 7+ (17 keV/nm) and 70 MeV Ni 5+ (11 keV/nm)) in three pyrochlores (Gd 2Zr 2O 7, Nd 2Zr 2O 7 and Gd 2Ti 2O 7) with progressively increasing radius ratios ( r A/ r B of A 2B 2O 7) of cations is reported. Since I +7 is one of the ions used, these measurements also simulate fission fragment damage in pyrochlores and identify a potential host lattice for inert matrix fuel. X-ray diffraction on the irradiated materials indicate amorphization in Gd 2Ti 2O 7 at the lowest S e used, and a transition to anion-deficient fluorite ( Fm 3 ¯ m ) structure with about 1% decrease in volume in Nd 2Zr 2O 7 and Gd 2Zr 2O 7 at higher S e. In Nd 2Zr 2O 7 this is followed by amorphization and Gd 2Zr 2O 7 does not amorphize even at the highest S e employed. Raman analysis of Gd 2Zr 2O 7 and Nd 2Zr 2O 7, indicate an increase of Zr coordination number after irradiation and the bands become broader due to disordering. Analysis of the results shows that the radiation susceptibility of these pyrochlores in the electronic stopping regime strongly depends on the radius ratio of A to B cations and hence on the energy required for formation of cation antisites and anion Frenkel pairs, similar to their susceptibility in the nuclear stopping regime.

Irradiation-enhanced interdiffusion in the diffusion zone of U-Mo dispersion fuel in Al

Uranium-molybdenum (U-Mo) alloy fuel particles dispersed in an aluminum (Al) matrix, designated as U-Mo/Al dispersion fuel, is in the development stage in the worldwide RERTR (Reduced Enrichment for Research and Test Reactors) program. The main issue in developing U-Mo/Al dispersion fuel is the diffusion reaction occurring at the interface between the fuel particles and matrix. To accurately analyze fuel performance, a model to predict the diffusion kinetics is necessary. For this purpose, the authors developed a diffusion layer growth rate correlation for out-of-pile annealing tests and a similar correlation for in-reactor tests. The correlation for in-reactor tests is considerably different from that of out-of-pile tests because it contains factors that amplify diffusion kinetics by fission damage in the diffusion reaction zone. This irradiation enhancement was formulated by a combination of the fission rate in the fuel and fission fragment damage distribution in the diffusion reaction zone. Using a computer code, fission damage factors were obtained as a function of diffusion reaction layer thickness and composition. The model correlation was established and fitted to the in-reactor data. As a result of this data fitting, the interaction layer growth rate is found to be proportional to the square root of the fission fragment damage rate and to have a temperature dependence characterized by the effective activation energy of 46 to 76 kJ/mole, which is smaller by a factor of 4 to 7 than that of out-of-pile tests.

Use of ion beams to simulate reaction of reactor fuels with their cladding

Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium–molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium–molybdenum coated with a 100nm thin film of aluminum was irradiated with 3MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27Al(p,γ)28Si reaction which occurs at a proton energy of 991.9keV. During irradiation at 150°C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30nm2/dpa. At a dose of 90dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery.

Nanoscale occurrence of Pb in an Archean zircon

We report, for the first time, a direct, atomic-scale characterization of Pb in zircon (4.4–3.1 Ga) from the early Archean Yilgarn craton in Australia using high-resolution HAADF-STEM. Two forms of Pb have been identified: Pb concentrated at ∼3 atom% as a nanoscale patch in zircon structure, and Pb concentrated within the amorphous domain created by fission fragment damage. The first result suggests that the Pb atoms directly substitute for Zr 4+ in the zircon structure, and the latter observation demonstrates that Pb diffusion can occur through amorphous regions created by radiation damage, although volume diffusion is typically considered to be the dominant mechanism for Pb diffusion. Beyond the first percolation point, i.e., when the amorphous domains overlap and form a fully interconnected network of amorphous domains, there is a new pathway for the diffusion of Pb that is faster than volume diffusion through crystalline zircon.

Effects of fission-fragment damage on vortex dimensionality in silver-sheathedBi2Sr2Ca2Cu3Oxtapes

We report on the vortex dimensionality of uranium-doped Ag/Bi2223 tapes, before and after irradiation to a thermal-neutron fluence. The effective activation energies as a function of current density and applied field were calculated from dynamic magnetisation relaxation measurements. A dimensional crossover from a 3D elastic creep regime to 2D plastic creep was observed in the nonirradiated tape at an applied magnetic field Hcr = 0.37 T, with an associated change in the flux hop velocity and temperature dependence of the activation energy. After introduction of the fission-fragment damage by irradiation, a shift in the crossover to Hcr = 0.65 T was observed. These results indicate an enhancement of the c-axis vortex coherence by the introduced splayed columnar defects, explaining the greater pinning efficiency of the uranium-fission method in Bi2223 rather than the less anisotropic Y123. Conflicting results obtained for the irradiated tape in the absence of any temperature scaling of the activation energies demonstrate the importance of the inclusion of scaling in the magnetisation relaxation analysis.

Silicon carbide as an inert-matrix for a thermal reactor fuel

This paper reports progress on work to develop methods of fabricating silicon carbide with cerium, as a substitute for plutonium, to achieve high densities at low sintering temperatures. Densities of 97–99% of TD were achieved at 1943 K for cerium oxide concentrations in the starting powders from 5 to 20 wt%. Also reported are: specific heat and thermal conductivity measurements of as-fabricated SiC; compatibility of SiC with coolant and Zircaloy-4; and accelerator simulations of in-reactor fission-fragment damage. The thermal conductivity for as-fabricated SiC with additives was 48 W m −1 K −1 at 298 K decreasing to about 18 W m −1 K −1 at 1773 K. Calculations, based on the measured thermal conductivity, show that the inert-matrix fuel could operate at 55 kW m −1 linear power at a centre-line temperature of only 673 K, i.e., only 100 K above coolant temperature, although it is expected that irradiation-induced degradation of thermal conductivity will lead to higher operating temperatures as burnup accumulates. The increase in central temperatures due to a possible decrease in thermal conductivity caused by fast-neutrons are calculated in the text. SiC appears to be a very promising candidate as an inert-matrix fuel for water-cooled reactors.

Isotropic short columnar pinning centers from fission fragment damage in bulk melt-textured YBCO

Uranium, with a controllable amount of U235, is admixed with YBCO, melt-textured, and irradiated with thermal neutrons, to fission U235. Fission fragments produce short (∼20μm), isotropic columnar damage, which act as pinning centers. In samples 2cm in diameter and 0.8cm thick, Jc of 65000 A/cm2 and trapped fields ∼2.1T at 77K have been achieved. A mini-magnet composed of four such samples traps 3.2T at 77K, and 8.1T at 62K.

High-resolution TEM study of ion beam irradiation induced amorphization in ceramic materials

Radiation damage of nuclear materials (e.g. fast- or fusion-neutron damage in reactor structural components, fission-fragment damage in nuclear fuels and alpha decay damage in nuclear waste forms) has been one of the major challenges faced by the material science community. Ion beam irradiation and implantation experiments have been used extensively in the past few decades not only for simulating these damaging process in materials but also for improving material properties for many technological applications.As an energetic particle traverses a crystalline target, it loses its energy predominantly through electronic (ionization) and nuclear (elastic collision) interactions with the atoms in the lattice. The target atom which receives sufficient energy from the interactions may get displaced from its lattice site and may further displace other target atoms, thus creating a displacement cascade which is usually a few nanometers in scale. Just at the end of the collision phase, which lasts for only a few tenths of a picosecond, a displacement cascade contains a very dense cluster of point defects and the region may be considered amorphous.

Effect of internal fission-fragment irradiation on critical currents and flux creep in Bi–Sr–Ca–Cu–O superconductors doped with UO2

Fission fragment damage was introduced into samples of Bi2Sr2CaCu2Ox and Bi1.7Pb0.3Sr2Ca2Cu3Oy, doped with various levels of UO2, by irradiation with thermal neutrons. The critical temperatures were unchanged. Concurrent with an increase in intragranular Jc previously reported, a decrease in flux creep was observed. The apparent pinning potential for creep at 10 K and 0.8 T increased on irradiation by about two to three times for both the 2212 and 2223 compounds. This increased apparent pinning potential is attributed to the strong pinning introduced by the damage caused by the travel of the fission fragments through the crystal. Pinning potential after irradiation increased with an increase in the amount of UO2 in the sample. The increase in bulk pinning potential on irradiation was proportional to the increase in intragranular critical currents on irradiation, qualitatively as expected theoretically.

Flux creep after thermal neutron irradiation of uranium-doped oriented compacts of YBa2Cu3Ox

Fission fragment damage was introduced by thermal neutron irradiation into aligned, sintered powder samples of YBa2Cu3Ox with uranium additions of 0.08 wt % UO2. This resulted in an increase in the intragranular critical current and a decrease in flux creep. Flux creep of the magnetization M exhibited simple log time dependence and was characterized by absolute and normalized slopes as functions of magnetic field and temperature. Before irradiation, the slopes were strongly field dependent and showed peaks as a function of temperature. After irradiation the slopes were much less field dependent and the peaks were suppressed. Both the magnitude and the field dependence of the creep were improved. Apparent pinning potentials at 0.8 T increased by a factor of about 2 on irradiation, for temperatures from 10 to 35 K. These increases in the apparent pinning potentials are attributed to the damage caused by the travel of the fission fragments through the crystal.

Flux pinning and flux creep in uranium-doped (Bi, Pb)-Sr-Ca-Cu-O superconducting powders after thermal-neutron irradiation

Fission fragment damage was introduced into uranium-doped powdered Bi/sub 1.7/Pb/sub 0.3/Sr/sub 2/Ca/sub 2/Cu/sub 3/ O/sub y/ by irradiation with thermal neutrons. Measurements of magnetic hysteresis and flux creep were made using a vibrating sample magnetometer. Magnetic hysteresis and intragranular critical current densities, obtained from the magnetic hysteresis using the critical state model, show an increase upon irradiation of a factor of 70 at 50 K and 0.8 T. Flux creep data, interpreted as nonlinear relations between pinning energy and magnetization, as temperature-dependent pinning energies, or as distributions of pinning energies, show significant increases in pinning energy upon irradiation. The irreversibility line is found to shift to higher magnetic fields upon irradiation. It is concluded that the increase in flux creep and decrease in critical current density at higher temperatures appear to limit the potential utility of this family of oxide superconductors to the lower-temperature regime.