Rock-like oxide fuel behavior under reactivity initiated accident conditions

Pulse irradiation tests of two types of rock-like oxide (ROX) fuel, i.e. yttria stabilized zirconia (YSZ) and YSZ/Spinel composite, were conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under reactivity-initiated accident conditions. The ROX fuels failed with cladding burst at fuel volumetric enthalpies above 10 GJ m⁻³, which was comparable to that of UO₂ fuel. The failure of the ROX fuels, however, occurred with considerable fuel melting and was quite different to that of UO₂ fuel, which was caused by cladding melting and embrittlement due to heavy oxidation. Lower fuel melting temperature of the ROX fuels compared to that of UO₂ contributed to the different fuel failure modes. Certain amount of molten ROX fuel dispersed out at the failure. However, the mechanical energy generation due to the molten fuel/water interaction was negligible for the ROX fuels at peak fuel enthalpies below 12 GJ m⁻³.     

Post irradiation examination of uranium inclusions in inert matrices.

The inert matrix materials CeO₂, MgO, Y₂O₃, MgAl₂O₄ and Y₃Al₅O₁₂ were selected as candidates for inert matrices for the EFTTRA²-T3 neutron irradiation experiment. Most targets contain 20% enriched ²³⁵U fissile inclusions with an average size of roughly 150 μm. The volume fraction of the fissile phase is either 2.5 vol% UO₂ or 19.6 vol% of Y_5.78UO_x in the inert matrices. The samples were irradiated for 198.9 full power days in the High Flux Reactor in Petten. The calculated burn-up is between 17.3 and 19.5% FIMA. The temperature of the cladding was kept at 600 ± 25 K. A dimensional change of at least +5 % is measured after neutron irradiation for Y₃Al₅O₁₂ and MgAl₂O₄ with macro dispersions of UO₂; the other targets with a macro dispersion of UO₂ show a volume change of less than + 1 vol%. The fractional release of the fission gas Xe is more than 40% for the MgAl₂O₄ and Y₃Al₅O₁₂ matrices with a macro dispersion of UO₂, the other targets show a fractional release of Xe of less than 15%. Cracks are observed in MgO and MgAl₂O₄ targets which is possibly related to the stress caused by swelling of the UO₂ inclusions.     

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl₂O₄ spinel and ZrO₂ are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl₂O₄ as an inert matrix, 2 targets contain ZrO₂ and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm⁻³. Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.     

Order–disorder phase transition induced by swift ions in MgAl2O4 and ZnAl2O4 spinels

Grazing X-ray diffraction was used to study in details the behaviour of two spinels, ZnAl₂O₄ and MgAl₂O₄ irradiated by swift ions. Such an irradiation allows to have an important irradiated depth and then accurate diffraction diagrams. Rietveld refinement done on these diagrams clearly show an order–disorder transition due to a melting of cations under irradiation in these two spinels. More especially, these results clearly exhibit that no new phase is created under irradiation in MgAl₂O₄. Raman spectroscopy, sensitive to the crystallographic space group, seems to confirm this analysis.     

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.     

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.     

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

The radiation stability of spinel MgAl₂O₄ 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.     

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.     

Radiation effects on MgAl2O4–yttria stabilized ZrO2 composite material irradiated with Ne ions at high temperatures

Spinel (MgAl₂O₄) and yttria stabilized ZrO₂ (YSZ) are candidates for fuel materials for use in nuclear reactors and the optical and insulating materials for fusion reactors. In our previous studies, the amorphization of spinel under 60 keV Xe ion irradiation at RT was observed. On the other hand, amorphization could not be confirmed in YSZ single crystals under the same irradiation conditions. In the present study, the damage evolution process of polycrystalline spinel–YSZ composite materials has been studied by in situ TEM observation during ion irradiation. The irradiation was performed with 30 keV Ne⁺ ions at a flux of 5 × 10¹³ ions cm⁻² s⁻¹ at 923 K and 1473 K, respectively. The observed results revealed a clear difference in morphology of damage depending on irradiation temperature and crystal grains. In the irradiation at 923 K, defect clusters and bubbles were formed homogeneously in YSZ grains. On the other hand, at 1473 K, only bubble formation was observed. The bubbles grew remarkably with increasing ion fluence in both grains. Even though the growth of the bubbles was observed in both grains, the average diameter of grown bubbles in spinel grains was larger than those in YSZ ones. The bubbles tended to form along the grain boundary at both temperatures.     

Surface modification of MgAl2O4 and oxides with heavy ions of fission fragments energy

In this paper we report on results of surface modification in several candidate materials for inert matrix fuel hosts (MgAl₂O₄, MgO and Al₂O₃) induced by (0.5–5) MeV/amu Kr, Xe and Bi ion bombardment in the fluence range of 2 × 10¹⁰–10¹² ions/cm². The size and shape of nanoscale hillock-like defects, each of which was created by the impact of a single ion, have been studied by using atomic force microscopy (AFM) technique. It was found that mean hillock height on sapphire and spinel surfaces depends linearly on the incident electronic stopping power. The hillocks are highest in MgAl₂O₄, having a lower threshold for the lattice disorder in the bulk material via relaxation of electronic excitations. As a possible reason for the hillocks formation, the plastic deformation due to the defects created by the Coulomb explosion mechanism in the target subsurface layer is suggested.     

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.     

氯化锂熔盐中八氧化三铀电解还原基础研究

将氧化物乏燃料直接电解还原为粗金属的过程是目前以电解还原-电解精炼为特征的主流干法后处理流程的重要步骤。二氧化铀（UO₂）是乏燃料的最主要成分,将致密的UO₂芯块转化为八氧化三铀（U₃O₈）粉末后,再进行电化学还原能有效提高还原速率。因此,以U₃O₈为研究对象,开展其在氯化锂（LiCl）熔盐中的电解还原机理研究,对后处理干法流程的开发具有重要的现实意义。本文在650 ℃的LiCl熔盐中,采用循环伏安法和恒电位电解法,研究U₃O₈的电解还原行为;对电解后的样品,运用XRD、SEM等手段分析其组成和形貌,并推测相应的还原机理。     

Melting behaviour of oxide systems for heterogeneous transmutation of actinides. II. The system MgO–Al2O3–PuO2

Isothermal sections of the phase diagram of the system MgO–Al₂O₃–PuO₂ at various temperatures were calculated using sublattice models. The results show that below 2133 K no liquid occurs in the system. Above 2133 K liquid starts to form at the Al₂O₃–PuO₂ side. The phase diagram of the pseudo-binary system PuO₂–MgAl₂O₄ was also obtained from an isopleth T–x calculation.     

Design of a gadolinia bearing mixed-oxide fuel assembly for pressurized water reactors

A study on neutronics design of a gadolinia (Gd₂O₃) bearing mixed-oxide (MOX) fuel assembly (MOX-UO₂ (Gd₂O₃) assembly) was performed for the purpose of suppressing the use of fresh lumped burnable poison rods (BPRs). The MOX-UO₂ (Gd₂O₃) assembly investigated consists of MOX and UO₂ (Gd₂O₃) fuel rods, which have already been verified through both fabrication and irradiation experiences. In all, 16 UO₂ (10 wt% Gd₂O₃) fuel rods are located at every corner and the peripheral region of the MOX-UO₂ (Gd₂O₃) assembly in order to reduce the power peaking of MOX fuel rods due to the thermal neutron inflow, and to reduce the reactivity penalty at the end of cycle (EOC). Since fresh BPRs are not expected to be inserted and UO₂ (Gd₂O₃) fuel rods are located at every corner of the assembly, the number of splits in plutonium (Pu) content can be only two, which is less than three splits required for a standard MOX assembly. Core characteristics of an equilibrium core loaded with MOX-UO₂ (Gd₂O₃) assemblies are evaluated and it is verified that adoption of the MOX-UO₂ (Gd₂O₃) assembly is effective to avoid the use of fresh BPRs with securing both the core safety and cycle length. The simplication of the splits in Pu content is also supposed to be beneficial, since it has the possibility of reduce MOX fuel fabrication costs.     

Structure and kinetic studies of U(VI)-benzamidoxime complex in non-aqueous solutions by H- and C-NMR

The structural and kinetic studies of U(VI) complex with benzamidoxime(Hba) as ligand in CD₃COCD₃ have been studied by means of ¹H and ¹³C NMR. The Hba molecule was found to coordinate to UO₂²⁺ in the form of anionic benzamidoximate (ba), and the number of ba coordinated to UO₂²⁺ was determined to be 3 by analyzing the chemical shift of ¹³C NMR signal for Hba in the presence of UO₂²⁺. The exchange rate constants(k_ex) of ba in [UO₂(ba)₃] were determined by the NMR line-broadening method. The kinetic parameters were obtained as follows: k_ex(25°C) = 3.1 × 10³s⁻¹, ΔH^≠ = 35.8 ± 3.5 KJ mol⁻¹, and ΔS^≠ = −65 ± 13.7 J K⁻¹ mol⁻¹. The UV-visible absorption spectra of solutions containing UO₂²⁺ and Hba were also measured. The molar extinction coefficient of the complex was found to be extremely large compared with those of UO₂(L)₅²⁺ (L = unidentate oxygen donor ligands) complexes. This is due to the strong electron withdrawing of UO₂²⁺ from Hba and suggests that an interaction between UO₂²⁺ and Hba is very strong. Such a high affinity of monomeric amidoxime to UO₂²⁺ reasonably explains the high adsorptibility of amidoxime resin to U(VI) species, and is considered to result in the high recovery of U(VI) species from sea water using amidoxime resin.     

Argon irradiation damage at a Cu/Al2O3 interface for different alumina structures

The evolution of damages at a Cu/Al₂O₃ device interface after Ar⁺ irradiation, depending on alumina structure, and the effect of surface roughness on sputtering have been studied. A polycrystalline Cu/Al₂O₃ bilayer and polycrystalline Cu on amorphous alumina were irradiated with 400 keV Ar⁺ ion beam at doses ranging from 5 × 10¹⁶ to 10¹⁷ Ar⁺/cm² at room temperature. The copper layer thicknesses were between 100 and 200 nm. RBS analysis was used to characterize the interface modification and to deduce the sputtering yield of copper. The SEM technique was used to control the surface topography. A RBS computer simulation program was used to reproduce experimental spectra and to follow the concentration profile evolutions of different elements before and after ion irradiation. A modified TRIM calculation program which takes into account the sputtering yield evolution as well as the concentration variation versus dose gives a satisfactory reproduction of the experimental argon distribution. The surface roughness effect on sputtering and the alumina structure influence at the interface on mixing mechanisms are discussed.     

含钆堆芯平衡循环优化技术研究

通过对平衡循环燃料管理技术进行分析,确定对钆棒采用两端装UO₂芯块的轴向分区设计并提高UO₂-Gd₂O₃芯块中235U富集度的优化方法。根据钆棒两端不同长度UO₂芯块对堆芯轴向功率分布的影响初步确定了钆棒两端UO₂芯块的长度,根据UO₂-Gd₂O₃芯块中235U富集度对燃料经济性和制造的影响初步确定了UO₂-Gd₂O₃芯块中的235U富集度。分析了钆棒轴向分区和提高UO₂-Gd₂O₃芯块中235U富集度各自及综合相对于比较基准方案对堆芯功率分布的影响,优化方案相对于比较基准方案在Ⅰ和Ⅱ类反应性事故工况下对安全性的影响,并对优化方案中的UO₂-Gd₂O₃芯块进行了安全验证。研究结果表明,通过在钆棒两...     

Fuel performance evaluation of rock-like oxide fuels

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ–spinel–corundum system, the single-phase fuels of YSZ, the particle-dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation testings. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO₂ fuel at the similar fuel temperatures. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.     

Structural study on uranyl ion in 1-butyl-3-methylimidazolium nonafluorobutanesulfonate ionic liquid

The structure of uranyl ion in 1-butyl-3-methylimidazolium nonafluorobutanesulfonate ionic liquid (BMINfO) has been studied with ¹H- and ³⁵Cl-NMR, Raman, and UV-visible spectroscopy. In the ¹H-NMR spectrum of the BMINfO solution prepared by dissolving UO₂(ClO₄)₂·5 6H₂O, the signal of H₂O coordinated to UO₂²⁺ was observed at 6.64 ppm at 50°C (free H₂O in BMINfO: 3.1 ppm at 50°C), suggesting that the uranyl species exists as the aquo complex, [UO₂(H₂O)_n]²⁺. The signal of the coordinated H₂O disappears with heating at 120°C for 3 h under vacuum. This indicates the dehydration from [UO₂(H₂O)_n]²⁺. On the other hand, the ³⁵Cl-NMR signal of ClO₄⁻ as the counter anion of UO₂²⁺ was observed at 1011 ppm (vs. Cl⁻ in D₂O) regardless of heating. This indicates that no ClO₄⁻ ion is in the first coordination sphere of UO₂²⁺. Furthermore, the UV-visible absorption spectra showed that the characteristic absorption bands due to UO₂²⁺ were sharpened with the dehydration. This means the simplification of the structure around UO₂²⁺. These results described above suggest that UO₂²⁺ in BMINfO has no ligand in its equatorial plane after the dehydration, i.e. UO₂²⁺ exists as a bare cation in this system.