Oxygen potential of high burn-up fast breeder oxide fuel

The oxygen potential of fast breeder mixed oxide fuel (U_0.8Pu_0.2)O_1.98 irradiated to different burn-ups up to 11 at% has been determined between 900 and 1300 K by measurements of the electromotive force in a galvanic microcell. The oxygen potential increases continuously with burn-up, due to the oxidative nature of fission and due to fission products dissolved in the fuel matrix. Lattice parameter measurements of similar fuel indicate that the fuel with the initial O/M ratio of 1.98 is still substoichiometric even at 7% burn-up. If the lattice parameter measurements are accepted, the increase in due to fission products is larger than assumed so far.     

The effect of fission products on the rate of U3O8 formation in SIMFUEL oxidized in air at 250°C

The effect of fission products on the rate of U₃O₈ formation was investigated by oxidizing UO₂-based SIMFUEL (simulated high burnup nuclear fuel) and unirradiated UO₂ fuel specimens in air at 250°C for different times (1–317 days). The progress of oxidation was monitored by X-ray diffraction, revealing that the rate of U₃O₈ formation declines with increasing burnup. An expression was derived to describe quantitatively the time for U₃O₈ powder formation as a function of simulated burnup. These findings were supported by additional isochronal oxidation experiments conducted between 200 and 300°C.     

Application of lattice constant measurements for stoichiometry determination in irradiated U-Pu mixed oxide fuels

A method is proposed for determining the oxygen/ metal ratio in mixed irradiated uranium-plutonium oxides. The method is based on a measurement of the lattice constants and on a standard thermal treatment which is used to obtain a tetravalent state of uranium and plutonium. The effects of irradiation and of the solution of solid fission products in the matrix, the variation in plutonium concentration, and influence of these factors on stoichiometry are discussed on the basis of the results of simulated experiments in which the state after irradiation of oxide fuels is computed, together with the concentrations of the fission products.

For a given burn-up τ the oxygen/metal ratio of the matrix O/U + Pu + FP, which has a considerable influence on the physical properties of the fuel, is obtained by direct measurement of the ratio O/U + Pu and correcting this value for the effect of soluble fission products using the equation: O/U + Pu + F.P. = 3/(3− τ) [(1− τ)O/U + Pu] + [2/3 τ·1.75].     

Neutron irradiation effects in uranium monosulfide

Uranium monosulfide (US) was irradiated to investigate the effects of fission damage. Post-irradiation examinations were done by measuring the electrical resistivity, and partly the magnetic properties, at low temperature. The lattice parameter and the electrical resistivity measured at room temperature just after the irradiations showed an increase starting at a fission dose of 1 × 10¹⁶ fissions/cm³ and attaining a maximum at 3 × 10¹⁶ fissions/cm³. After that, a saturation of both increases persiste until 3 × 10¹⁷ fissions/cm³. The low-temperature electrical resistivity in the magnetic ordered state (ferromagnetic transition, T_c, at about 180 K) increased remarkably, while decreasing drastically in the magnetization, with increasing fission dose, apparently corresponding to the lattice expansion. In addition, the Curie point (T_c) shifted to lower temperatures with accumulating fission damage.     

Analysis of Cu and O in high-Tc superconductors by 7–9 MeV protons

The use of proton-induced γ-ray emission for the simultaneous analysis of Cu and O in high-T_c superconductors is demonstrated. Utilizing 7–9 MeV protons, the ratio of O relative to Cu can be determined reliably to a few percent accuracy in homogeneous bulk samples and films thicker than 130 μm using standard bulk samples of O and Cu. Results of the present method are compared with those of the analysis of YBa₂Cu₃O_7xdone by the Rutherford scattering of particles and the non-Rutherford scattering through the ¹⁶O(p,p₀)¹⁶O reaction.     

Measurement of the mechanical properties of thermally-shocked zirconia-based simulated inert matrix fuel

The Vickers micro-hardness (H_V) was measured by an indentation technique of simulated ZrO₂-based Inert Matrix Fuel (IMF) material with a composition of Er_0.07Y_0.10Ce_0.15Zr_0.68O_1.915 in two different densities on sintered specimens and specimens thermally shocked with the quenching temperature differences (ΔTs) between 473 and 1673 K and compared with those of simulated MOX, namely, U_0.92Ce_0.08O₂. The H_V values obtained for two IMF materials were found higher, ranging from 6.37 GPa to about 7.84 GPa, depending on ΔT and the sintered density, than those obtained for the simulated MOX which are quasi-constant in the same range of ΔT with a mean value of 6.37 GPa. The fracture toughness (K_IC) was calculated from the measured H_V and the crack length, and it was found to exhibit a slight increase with increasing ΔT, ranging between 1.4 and 2.0 MPa m^1/2, while that of simulated MOX specimen is ranging between 0.8 and 1.1 MPa m^1/2. The thermally shocked specimens were observed with an optical microscope and analyzed in terms of microstructural changes and cracking patterns.     

Observation of a high burnup rim-type structure in an advanced plutonium–uranium carbide fuel

The observation is reported of a ‘rim-type' structure with small subgrains in an advanced plutonium–uranium carbide (U_0.8Pu_0.2)C fuel pin, which had been irradiated in the Dounreay Fast Reactor to a burnup of 8.3% FIMA.     

Evaluation on chemical state of irradiated rock-like oxide fuels by nuclear and chemical modeling

Chemical forms of fission products in irradiated ROX fuels were calculated by the SOLGASMX-PV code, and the resultant phase equilibrium and the oxygen potential in the fuel were evaluated in order to assess the irradiation behavior of the ROX fuels. For the ROX fuel with reactor grade Pu, the oxygen potential increased to about −140 kJ mol⁻¹ at EOL when all the Pu in the fresh fuel was tetravalent. In the case of fresh fuel which was partially reduced with the [Pu⁺³]/[Pu⁺⁴]=10/90, the oxygen potential increase was suppressed to about −400 kJ mol⁻¹. On the other hand, the oxygen potential of the ROX fuel with weapon grade Pu never exceeded the value of about −400 kJ mol⁻¹. The difference of oxygen potentials was caused by difference of Am amount produced by Pu conversion. The oxygen potential of the irradiated fuel was controlled by the phase equilibria among FPs. The equilibrium between metallic Mo and MoO₂ controlled the oxygen potential to about −400 kJ mol⁻¹.     

Irradiation effects of swift heavy ions in actinide oxides and actinide nitrides: Structure and optical properties

Actinide oxides have been used as nuclear fuels in the majority of power reactors working in the world and actinide nitrides are under investigation for the fuels of the future fast neutron fission reactors developed in Forum Generation IV. Radiation damage in actinide oxides UO₂, (U_0.92Ce_0.08)O₂, and actinide nitride UN has been characterized after irradiation with swift heavy ions. Fluences up to 3 × 10¹³ ions/cm² of heavy ions (Kr 740 Mev, Cd 1 GeV) available at the CIRIL/GANIL facility were used to simulate irradiation in reactors by fission products and by neutrons. The macroscopic effects of irradiation remains very weak compared with those seen in other ceramic oxides irradiated in the same conditions: practically no swelling can be measured and no change in colour can be observed on the irradiated part of a polished face of sintered disks. The point defects in irradiated actinide compounds have been characterized by optical absorption spectroscopy in the UV–Vis–NIR wavelength range. The absorption spectra before and after irradiation are compared, and unexpected stability of optical properties during irradiation is shown. This result confirms the low rate of formation of point defects in actinide oxides and actinide nitrides under irradiation. Actinide oxides and nitrides studied are >40% ionic, and oxidation state of the actinides seems to be stable during irradiation. The small amount of point defects produced by radiation (<10¹⁶ cm⁻²) has been identified from differences between the absorption spectrum before irradiation and the one after irradiation: point defects in oxygen or nitrogen lattices can be observed respectively in oxides and nitrides (F centres), and small amounts of U⁵⁺ would be present in all compounds.     

Formation of uranium mononitride by the reaction of uranium dioxide with carbon in ammonia and a mixture of hydrogen and nitrogen: II. Reaction rates

Kinetics of the carbothermic synthesis of UN from UO₂ in an NH₃ stream and a mixed 75% H₂ + 25% N₂ stream were studied in the temperature range of 1400–1600°C by X-ray analysis and weight change measurement of the sample. The weight change was divided into two parts; i.e. weight loss due to carbothermic reduction of UO₂ and weight loss due to removal of carbon by hydrogen. The former followed the first-order rate equation −1n(1 − ₀) = k₀t, and the latter the rate equation of phase boundary reaction 1 − (1 − _c)^1/3 = k_ct. The apparent activation energy of the former was in the range of 320–380 kJ/mol. The value of the latter in an NH₃ stream was 175–185 kJ/mol, which was smaller than that in a mixed 75% H₂ + 25% N₂stream (285 kJ/mol). In this method, the rate of the removal of carbon by hydrogen determines that of the formation of high purity UN.     

基于多物理场耦合的U3Si2燃料与双层SiC包壳组合的轻水堆燃料性能分析

基于COMSOL平台开发了一套基于多物理场全耦合的燃料性能分析程序,并通过径向功率分布模型对比验证了该程序的正确性与准确性;然后进一步分析了U₃Si₂燃料与双层SiC包壳组合、U₃Si₂燃料与锆合金包壳组合在反应堆正常运行工况下的性能,并与UO₂燃料与锆合金的组合进行了对比分析。计算结果发现U₃Si₂燃料与锆合金包壳组合相比UO₂燃料与锆合金的组合具有更低的燃料中心温度、裂变气体释放量及内压,但气隙闭合时间会提前;而U₃Si₂燃料与双层SiC包壳的组合相比U₃Si₂燃料与锆合金的组合具有更高的燃料中心温度、更大的裂变气体释放量及内压,且随着燃耗的增加,其燃料中心温度大幅增加,与锆合金包壳相比,双层SiC包壳能够有效延迟气隙闭合,缓解燃料与包壳的力学相互作用。      

A mechanism for the ease of slip in UO2+x

The microstructure of plastically deformed hyperstoichiometric uranium dioxide crystals has been reexamined by transmission electron microscopy. Although the U₄O₉ phase is present with its usual superlattice structure, dislocations which thread interfaces between it and the adjacent UO₂ phase exhibit uniform contrast, (i.e., no separation into superdislocations is resolved). Dislocations lying totally within the UO₂ phase are decorated by “shadows” of gray contrast, which can be removed by heating in the microscope atmosphere. These have been identified as “atmospheres” of U₄O₉. A geometrical argument, based on the coordinates of anion interstitials published by Willis, is used to explain how these “atmospheres” might account for the observed incidence of extensive cross-slip in plastically deforming UO_2+x.     

Electrical conductivity and non-stoichiometry in the (U,Gd)O2±x system

The isothermal electrical conductivity and oxygen potential of the (U,Gd)O_2±x solid solution were measured in various oxygen partial pressure regions at 1200 °C and 1300 °C. The electrical conductivity gradually decreased with decreasing oxygen partial pressure even in the hypo-stoichiometric region. These findings were in contrast to the implication of a hypo-stoichiometry where the electrical conductivity is increased through the formation of oxygen vacancies. The (U_1−yGd_y)O_2−y/2 was defined as a new stoichiometric composition to determine the relationship between the deviation of the oxygen composition from stoichiometry and oxygen partial pressure. The dependence of the new oxygen deviation, z in (U_1−yGd_y)O_2−y/2+z, on the oxygen partial pressure corresponds to the dependence of the electrical conductivity, and thus a consistent defect structure model can be deduced from both the dependence curves. It suggests that the defect type is oxygen interstitial even below the oxygen composition of 2.     

Recommended values of the delayed neutron yield for: U-235; U-238 and Pu-239

Today there is no well-established theoretical model to predict the fission delayed neutron yield v_d with the required accuracy. In this field the recommended data result from the rare experimental data analysis or from purely phenomenological or semi-phenomenological models. There is another source of valuable information: the related integral data or β_eff- data. In this report we demonstrate, via a careful analysis of the experimental methods leading to revisited experimental β_eff values and associated uncertainties, that for the major nuclei the v_d evaluated data are of acceptable quality. For U-235 U-238 and Pu-239 we recommend v_d values for the thermal and the fast reactor ranges which have been obtained from a statistical consistent adjustment to the β_eff data. In the course of this study we show that the energy dependence of v_d, suspected from a physics point of view, probably exists with a different magnitude according to the nucleus. Concerning the major nuclei it is of negligible importance for the applications. The improvement of the higher Pu isotopes and minor actinides data is the main motivation for developing the theoretical investigations of the delayed neutron generation mechanism at the same level as the necessary experimental activity.     

RBS investigations of buffer layers between YBa2Cu3O7−x and silicon

The deposition of high-quality high-T_c superconducting films on silicon wafers for future hybrid electronic devices is strongly hampered by the interdiffusion between films and substrate. This effect degrades the superconducting properties seriously and is a strong function of temperature. Since high processing temperatures are inevitable for good films, suitable buffer layers are needed to reduce the interdiffusion. We have investigated the combinations ZrO₂/Si(100), BaF₂/Si(100), and noble-metal/TiN/Si(100) at temperatures up to 780°C in oxidizing ambient. YBa₂Cu₃O_7−x films have been deposited onto the buffer layers by laser ablation. Thereafter the interfaces have been analyzed by Rutherford backscattering. So far only ZrO₂ has demonstrated sufficient stability to serve as a buffer layer for the laser-ablated YBa₂Cu₃O_7−x films. All other combinations suffer from interdiffusion or oxidation.     

The constitution of the uranium-tin-palladium system

The Pd-rich region of the isothermal section of the ternary U---Sn---Pd system at 1050°C was investigated by metallography, X-ray microanalysis and X-ray diffraction. The fcc -Pd(Sn, U) solid solution dissolves up to 16 at.% Sn and 15 at.% U. The AuCu₃ type phases SnPd₃ and UPd₄ are the end members of a single-phase region. SnPd₂ and UPd₃ are in equilibrium with this solid solution of the composition U_0.68Sn_0.32Pd₃ and with the ternary phase USnPd₂. The Gibbs energies of formation of SnPd₂, SnPd₃ and UPd₄ were used and ideal behaviour of the (Sn, U) Pd_3+x solid solution was assumed to calculate the Gibbs energy of formation of UPd₃ which gives Δ_fG° = −312 kJ/mol at 1050°C. In addition, the annealing experiments in the Pd-rich region of the binary U---Pd system were extended to 950°C which confirm the phase-field distribution established at 1050°C in earlier work.     

Fission product precipitates in irradiated uranium carbonitride fuel

Austenitic steel-cladded uranium carbonitride fuel pins were irradiated in the BR2 up to 6.4% burnup. A cross-section of the pin RV 24 with the fuel composition UC_0.86N_0.09O_0.05 was prepared for X-ray microanalysis of the fission product precipitates. Rare-earth oxide and U(Mo,Tc)C₂ phases were observed in the whole fuel region. Bright phases present in annular rings of the outer fuel zone were identified as U₂(Tc, Ru, Rh)C₂. Alkaline-earth oxide and U–Pd–Ni phases were shown in the fuel-cladding gap. The rare-earth and alkaline-earth fission products extracted the oxygen from the fuel matrix which became nearly oxygen free. The formation of nitrides could not be detected.     

Phase transitions in U3O8−z: I, heat capacity measurements

The heat capacity of U₃O_8−z with various O/U ratios was measured in the range from 250 to 750 K, and λ-type heat capacity anomalies were found in each sample. The transition temperatures were 487 and 573 K for UO_2.663, 490 and 576 K for UO_2.656 and 508, 562 and 618 K for UO_2.640. The entropy changes of the transitions were 0.44 and 0.39 J K⁻¹mol⁻¹ for UO_2.663, 0.58 and 0.47 J K⁻¹mol⁻¹ for UO_2.656 and 0.62, 0.51 and 0.25 J K⁻¹mol⁻¹ for UO_2.640, increasing as O/U decreases. The enthalpy change due to the transition varied linearly with the transition temperature except for UO_2.640, showing the presence of the same mechanism of phase transition among the samples with various O/U ratios. The mechanism of the phase transition was discussed on the assumption that the transition is originated from the order-disorder rearrangement of U⁵⁺ and U⁶⁺ with a consequent displacement of atoms, similarly to the case of U₄O_9−y.     

Thermal removal of gallium from gallia-doped ceria

The potential for removing gallium from Ga-doped CeO_2−x by thermal means was studied for the purpose of assessing gallium removal from PuO_2−x. The latter is of interest to those considering the storage or use of weapon-grade Pu, for example as a mixed oxide fuel component. Experiments were done by varying temperature, gas composition, exposure time, sample size, particle size, and gas flow rate. The kinetics of gallium removal were assessed through measurements of weight change, scanning electron microscopy, and chemical analyses. Results suggest that the gallium level can be reduced significantly by thermal treatments in Ar–6% H₂. Gallium species segregate to grain boundaries because of the low solubility of Ga₂O₃ in CeO₂. The kinetics and microstructural observations suggest that both gaseous and solid-state diffusion of gallium species are important for the removal of gallium.     

Analysis of wall conditioning by using plasma impact desorption in the tandem mirror GAMMA 10

A method for evaluating wall condition by using plasma impact desorption (PID) technique has been developed and successfully applied to the tandem mirror GAMMA 10 as a monitor for wall conditioning. A magnetically shielded quadrupole mass spectrometer was installed in the vacuum chamber of the GAMMA 10 central cell. The behaviour of the partial pressure of various gas molecules desorbed by ICRF-heated plasma discharges were analyzed. The predominant increase of the partial pressure due to PID (ΔP_PID) was hydrogen (M = 2) and a small amount of impurity as CO (M = 28), CH₃ (M = 15), H₂O (M = 18) and CO₂ (M = 44) was observed in the wall-conditioning discharges. The reduction of hydrogen pressure as well as ΔP_PID of the above impurities was observed with the progress of wall conditioning. This behavior has a good correlation with the increase of partial pressures due to electron-impact desorption measured at the same period. The relation between ΔP_PID and the charge-exchange flux was investigated.