Response of synthetic coffinite to energetic ion beam irradiation

Coffinite, USiO₄, is one of the two most abundant and important naturally occurring U⁴⁺ phases (the other is UO₂), and it is an alteration product of the UO₂ in spent nuclear fuel when in contact with silica-rich groundwater under reducing conditions. Despite its ubiquity, there are very limited data on the response of coffinite to radiation. Here, we present the results of the first systematic investigation of energetic ion beam irradiation (1 MeV Kr²⁺) of ultra-fine, synthetic coffinite (20-50 nm). In situ transmission electron microscopy (TEM) showed that the crystalline-to-amorphous transformation occurs at a relatively low dose, ∼0.27 displacements per atom (dpa) at room temperature. The critical temperature, T_c, above which coffinite cannot be amorphized, is low (∼608 K). Synthetic coffinite is more stable as compared with isostructural zircon (ZrSiO₄; T_c = 1000 K) and thorite (ThSiO₄; T_c above 1100 K) upon ion beam irradiation at elevated temperature, suggesting enhanced defect annealing behavior in nano-sized synthetic coffinite. Irradiation was found to decrease the temperature required to induce phase decomposition process in coffinite upon thermal annealing. A good correlation among the critical amorphization temperature, T_c, phase decomposition temperature, T_f, and the temperature range of the two-phase (ZrO₂ and SiO₂) co-existed region was identified.     

Developing uranium dicarbide-graphite porous materials for the SPES project

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO₂, U₃O₈ and UO₃), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U₃O₈ and UO₃ oxides to UO₂ before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC₂ and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.     

Clarification of high density electronic excitation effects on the microstructural evolution in UO2

In order to understand the properties of ion tracks and the microstructural evolution under accumulation of ion tracks in UO₂, 100 MeV Zr¹⁰⁺ and 210 MeV Xe¹⁴⁺ ions irradiation examinations have been done at a tandem accelerator facility of JAEA-Tokai, and it has been observed the microstructure by means of a transmission electron microscope (TEM) and a scanning electron microscope (SEM) in CRIEPI.Comparison of the diameter of ion tracks between UO₂ and CeO₂ under irradiation with 100 MeV Zr¹⁰⁺ and 210 MeV Xe¹⁴⁺ ions at room temperature clarify that the sensitivity on high density electronic excitation of UO₂ is much less than that of CeO₂. By the cross-sectional observation of UO₂ under irradiation with 210 MeV Xe¹⁴⁺ ions at 300 °C, elliptical changes of fabricated pores that exist till ∼6 μm depth and the formation of dislocations have been observed in the ion fluence over 5 × 10¹⁴ ions/cm². The drastic changes of surface morphology and inner structure in UO₂ indicate that the overlapping of ion tracks will cause the point defects, enhance the diffusion of point defects and dislocations, and form the sub-grains at relatively low temperature.     

The effects of alpha-radiolysis on UO2 dissolution determined from batch experiments with Pu-doped UO2

The effects of alpha dose-rate on UO₂ dissolution were investigated by performing dissolution experiments with ²³⁸Pu-doped UO₂ materials containing nominal alpha-activity levels of ∼1-100 Ci/kg UO₂ (actual levels 0.4-80 Ci/kg UO₂), in 0.1 M NaClO₄ and in 0.1 M NaClO₄ + 0.1 M carbonate. Dissolution rates increased less than 10-fold for an almost 100-fold increase in doping level and fall within the range of predictions of the Mixed Potential Model (a detailed mechanistic model for used fuel dissolution). Dissolution rates were lower in carbonate-free solutions and enrichment of ²³⁸Pu on the UO₂ surface was suggested in carbonate solutions. Effective G values, defined as the ratio of the total amount of U dissolved divided by the maximum possible amount of U dissolved by radiolytically produced H₂O₂, increased with decreasing doping levels. This suggests that the dissolution reaction at high dose rates is limited by the reaction rate between UO₂ and H₂O₂, but becomes increasingly limited by the rate of production of H₂O₂ at lower dose rates.     

Effect of Nd and Pr addition on the thermal and mechanical properties of (U, Ce)O2

The thermal conductivity, Young’s modulus, and hardness of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.01, 0.08, 0.12) were evaluated and the effect of Pr and Nd addition on the properties of (U, Ce)O₂ were studied. The polycrystalline high-density pellets were prepared with solid state reactions of UO₂, CeO₂, Pr₂O₃, and Nd₂O₃. We confirmed that all Ce, Pr, and Nd dissolved in UO₂ and formed solid solutions of (U, Ce, Pr, Nd)O₂. We revealed that the thermal conductivity of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.12) was up to 25% lower than that of x = 0.01 at room temperature. The Young’s modulus of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ decreased with x, whereas the hardness values were constant in the investigated x range.     

Thermodynamics of fission products in dispersion fuel designs - First-principles modeling of defect behavior in bulk and at interfaces

Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO₂ and UO₂ oxides, and the MgO/(U, Hf, Ce)O₂ interfaces have been carried out. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO₂. However, this trend is reversed or reduced for alkaline earth oxides with larger cation sizes. In the case of UO₂, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-site Coulomb repulsion between the localized 5f electrons. The fission product solution energies in bulk UO_2 ± x have been calculated as a function of non-stoichiometry x, and were compared to that in MgO. The solution energies of fission products in MgO are substantially higher than in UO_2 ± x, except for the case of Sr in hypostoichiometric UO₂. Due to size effects, the thermodynamic driving force of segregation for Xe and Cs from bulk MgO to the MgO/fluorite interface is strong. However, this driving force is comparatively weak for Sr.     

Synthesis and characterization of coffinite

Coffinite, USiO₄, has been produced by hydrothermal synthesis. The synthesis products, coffinite nanoparticles (50 nm in size) with UO₂ nanoparticles (a few nanometers), are always associated even if they are not always detected by XRD measurements. The formation of coffinite was shown to be very sensitive to several experimental parameters. The most important of these parameters are the pH, which must be in the range 8-9.5, the pressure, which must be below 50 bars, and the reaction conditions, which must be oxygen-free to maintain uranium in its tetravalent oxidation state. XRD and TEM reveal that tetragonal coffinite accounts for more than 90% of the final products while the by-products UO₂ and a Si-rich amorphous phase are also present. The structural formula of the obtained coffinite is close to USiO₄ as determined from EMPA (U_0.99±0.06Si_0.97±0.07O₄). XPS measurements show a peak chemical shift of the U-4f core levels by 1 eV toward higher binding energies in coffinite compared with stoichiometric UO₂. The U-4f_7/2 and U-4f_5/2 positions in coffinite are found to be near 380.8 ± 0.3 eV and 391.7 ± 0.3 eV, respectively.     

Effect of external gamma irradiation on dissolution of the spent UO2 fuel matrix

Leaching experiments were performed on UO₂ pellets doped with alpha-emitters (^238/239Pu) and on spent fuel, in the presence of an external gamma irradiation source (A⁶⁰Co = 260 Ci,  Gy h⁻¹). The effects of α, β, γ radiation, the fuel chemistry and the nature of the cover gas (aerated or Ar + 4%H₂) on water radiolysis and on oxidizing dissolution of the UO₂ matrix are quantified and discussed. For the doped UO₂ pellets, the nature of the cover gas clearly has a major role in the effect of gamma radiolysis. The uranium dissolution rate in an aerated medium is 83 mg m⁻² d⁻¹ compared with only 6 mg m⁻² d⁻¹ in Ar + 4%H₂. The rate drop is accompanied by a reduction of about four orders of magnitude in the hydrogen peroxide concentrations in the homogeneous solution. The uranium dissolution rates also underestimate the matrix alteration rate because of major precipitation phenomena at the UO₂ pellet surface. The presence of studtite in particular was demonstrated in aerated media; this is consistent with the measured H₂O₂ concentrations (1.2 × 10⁻⁴ mol L⁻¹). For spent fuel, the presence of fission products (Cs and Sr), matrix alteration tracers, allowed us to determine the alteration rates under external gamma irradiation. The fission product release rates were higher by a factor of 5-10 than those of the actinides (80-90% of the actinides precipitated on the surface of the fragments) and also depended to a large extent on the nature of the cover gas. No significant effect of the fuel chemistry compared with UO₂ was observed on uranium dissolution and H₂O₂ production in the presence of the ⁶⁰Co source in aerated conditions. Conversely, in Ar + 4%H₂ the fuel self-irradiation field cannot be disregarded since the H₂O₂ concentrations drop by only three orders of magnitude compared with UO₂.     

Recovery and restructuring induced by fission energy ions in high burnup nuclear fuel

In light water commercial reactors, extensive change of grain structure was found at high burnup ceramic fuels. The mechanism is driven by bombardment of fission energy fragments and studies were conducted by combining accelerator based experiments and computer-science. Specimen of CeO₂ was used as simulation material of fuel ceramics. With swift heavy ion (Xe) irradiation on CeO₂, with 210 MeV, change of valence charge and lattice deviation of cations were observed by XPS and XRD. Combined irradiations of Xe implantation and swift heavy ion irradiation successfully produced sub-micrometer sized sub-grains, similar as that observed in commercial fuels. Studying components of mechanism scenarios, with first principle calculations using the VASP code, we found stable hyper-stoichiometric defect structures of UO_2+x. Molecular dynamics studies revealed stability of Xe planar defects and also found rapid transport mode of oxygen-vacancy clusters.     

Damage evolution in Au-implanted Ho2Ti2O7 titanate pyrochlore

Damage evolution at room temperature in Ho₂Ti₂O₇ single crystals is studied under 1 MeV Au²⁺ ion irradiation by Rutherford backscattering spectroscopy along the 〈0 0 1〉 direction. For a better determination of ion-induced disorder profile, an iterative procedure and a Monte Carlo code (McChasy) were used to analyze ion channeling spectra. A disorder accumulation model, with contributions from the amorphous fraction and the crystalline disorder, is fit to the Ho damage accumulation data. The damage evolution behavior indicates that the relative disorder on the Ho sublattice follows a nonlinear dependence on dose and that defect-stimulated amorphization is the primary amorphization mechanism. Similar irradiation behavior previously was observed in Sm₂Ti₂O₇. A slower damage accumulation rate for Ho₂Ti₂O₇, as compared with damage evolution in Sm₂Ti₂O₇, is mainly attributed to a lower effective cross section for defect-stimulated amorphization.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

Estimation of oxygen potential for U-Pu-O mixed oxide fuel

A thermochemical analysis of the estimation of oxygen potential for mixed oxide fuel is presented. The MOX fuel is treated as a pseudo-quaternary solid solution of UO₂-U_cO_d-PuO₂-Pu_aO_b with b = 1.5a and d = 2.25c. The hypo-stoichiometric fuel is viewed as UO₂-(1/a)Pu_aO_b-PuO₂ pseudo-ternary system and the hyper-stoichiometric fuel is viewed as UO₂-(1/c)U_cO_d-PuO₂ pseudo-ternary system and the graphical representation is visualized as a ‘diamond plot’. The oxygen potential for MOX is determined either by PuO₂-Pu_aO_b equilibria or by the UO₂-U_cO_d equilibria subject to the constraint that the oxygen potential is equal both in Pu-O subsystem and in U-O subsystem. The choice of stoichiometric parameters a and c affects the relative contribution of configurational entropy to the oxygen potential which determines the shape of the ‘S’-type curve near exact stoichiometry. From the analysis of literature data, a = 2 and c = 8 was derived at for the ‘diamond plot’ model. Oxygen potentials for MOX fuel with 21, 28% and 44% plutonia was estimated using the ‘diamond plot’ model. A simplified nomograph for the oxygen potential of MOX is presented.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Eutectic crystallization in the FeO1.5-UO2+x-ZrO2 system

Results of the investigation of the FeO_1.5-UO_2+x-ZrO₂ system in air are presented. The eutectic position and the content of the phases crystallized at this point have been determined. The temperature and the composition of the ternary eutectic are 1323 ± 7 °С and 67.4 ± 1.0 FeO_1.5, 30.5 ± 1.0 UO_2+x, 2.1 ± 0.2 ZrO₂ mol.%, respectively. The solubilities of FeO_1.5 and ZrO₂ in the UO_2+x(FeO_1.5, ZrO₂) solid solution correspond to respectively 3.2 and 1.1 mol.%. The solubilities of UO₂ and ZrO₂ in FeO_1.5 are not significant. The existence of a solid solution on the basis of U(Zr)FeO₄ compound is found. The ZrO₂ solubility in this solid solution is 7.0 mol.%.     

Oxygen potential of hypo-stoichiometric Lu-doped UO2

Oxygen potentials of hypo-stoichiometric Lu-doped UO₂, (U_0.80Lu_0.20)O_2−x, were experimentally investigated by thermogravimetric analysis using H₂O/H₂ gas equilibria at 1173, 1273 and 1473 K. The oxygen potentials of (U,Lu)O_2−x were higher than those of other forms of rare earth-doped UO₂, specifically (U,Nd)O_2−x, (U,Gd)O_2−x, and (U,Er)O_2−x. Slope analyses for plots of oxygen potential versus deviation from stoichiometry indicated that (U_0.80Lu_0.20)O_2−x had a similar defect structure to that of the other forms of rare earth-doped UO₂. A relationship between the effective ionic radii and oxygen potentials was found for the hypo-stoichiometric rare earth-doped UO₂.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

Kinetic and thermodynamic studies of the dissolution of thoria-urania solid solutions

The dissolution of Th_1−xU_xO₂ was investigated through leaching experiments combined with X-ray photoelectron spectroscopy (XPS) and X-ray absorption spectroscopy (XAS) analyses. These experiments were performed in acidic and in oxidizing conditions (nitric solutions), for several compositions of solid solutions ranging from x = 0.24 to 0.81. Static sequential experiments in acidic media performed at room temperature confirmed that higher concentration of uranium in the solid solution leads to higher release of uranium in the leachate whatever the pH. The normalized dissolution rate in oxidizing media is increasing all the more the content of uranium is increases in the mixed oxide. While for Th enriched solids, kinetic parameters remain similar to that of ThO₂, in the case of uranium enriched solids, a drastic change is observed, and kinetic parameters are similar to that of UO₂ ones. For x > 0.50, the saturation is reached in the leachate after 100 days. XPS and EXAFS analysis on leached samples pointed out an oxidation of U(IV) at the surface for x < 0.5, and in the bulk for x > 0.5. Enrichment in Th is also observed at the surface of the solid, indicating the formation of a protective layer of hydrated thorium oxide, or hydroxide. Finally, the solubility product of secondary phase was determined. The values obtained are in good agreement with that of ThO₂, Th(OH)₄ and ThO₂, xH₂O reported in the literature.     

Thermal studies on fluorite type ZryU1−yO2 solid solutions

Solid state reactions of UO₂ and ZrO₂ in mild oxidizing condition followed by reduction at 1673 K showed enhanced solubility up to 35 mol% of zirconium in UO₂ forming cubic fluorite type Zr_yU_1−yO₂ solid solution. The lattice parameters and O/M (M = U + Zr) ratios of the solid solutions, Zr_yU_1−yO_2+x, prepared in different gas streams were investigated. The lattice parameters of these solid solutions were expressed as a linear equation of x and y: a₀ (nm) = 0.54704 − 0.021x - 0.030y. The oxidation of these solid solutions for 0.1 ? y ? 0.2 resulted in cubic phase MO_2+x up to700 K and single orthorhombic zirconium substituted α-U₃O₈ phase at 1000 K. The kinetics of oxidation of Zr_yU_1−yO₂ in air for y = 0-0.35 were also studied using thermogravimetry. The specific heat capacities of Zr_yU_1−yO₂ (y = 0-0.35) were measured using heat flux differential scanning calorimetry in the temperature range of 334-860 K.     

Effect of temperature on studtite stability: Thermogravimetry and differential scanning calorimetry investigations

The main objective of this work is the study of the influence of temperature on the stability of the uranyl peroxide tetrahydrate (UO₂O₂ · 4H₂O) studtite, which may form on the spent nuclear fuel surface as a secondary solid phase. Preliminary results on the synthesis of studtite in the laboratory at different temperatures have shown that the solid phases formed when mixing hydrogen peroxide and uranyl nitrate depends on temperature. Studtite is obtained at 298 K, meta-studtite (UO₂O₂ · 2H₂O) at 373 K, and meta-schoepite (UO₃ · nH₂O, with n < 2) at 423 K. Because of the temperature effect on the stability of uranyl peroxides, a thermogravimetric (TG) study of studtite has been performed. The main results obtained are that three transformations occur depending on temperature. At 403 K, studtite transforms to meta-studtite, at 504 K, meta-studtite transforms to meta-schoepite, and, finally, at 840 K, meta-schoepite transforms to U₃O₈. By means of the differential scanning calorimetry the molar enthalpies of the transformations occurring at 403 and 504 K have been determined to be −42 ± 10 and −46 ± 2 kJ mol⁻¹, respectively.     

Fabrication method for UO2 pellets with large grains or a single grain by sintering in air

The effects of a powder treatment, the sintering temperature and the sintering time on the grain growth of UO₂ pellets were investigated in air to obtain UO₂ pellets with large grains. Air could be used for sintering because an oxidation path above 1803 K does not pass through a two-phase (UO_2+x + U₃O_8−z) region. The UO₂ pellets sintered by the CO₂-air-CO₂-H₂ process consisted of a single grain or some large grains in the order of several millimeters.