Mechanical properties and XRD studies of silicon carbide inert matrix fuel fabricated by a low temperature polymer precursor route

The mechanical properties of silicon carbide (SiC) inert matrix fuel (IMF) pellets fabricated by a low temperature (1050 °C) polymer precursor route were evaluated at room temperature. The Vickers hardness was mainly related to the chemical bonding strength between the amorphous SiC phase and the β-SiC particles. The biaxial fracture strength with pre-notch and fracture toughness were found to be mostly controlled by the pellet density. The maximum Vickers hardness, biaxial fracture strength with pre-notch and fracture toughness achieved were 5.6 GPa, 201 MPa and 2.9 MPa m^1/2 respectively. These values appear to be superior to the reference MOX or UO₂ fuels. Excellent thermal shock resistance for the fabricated SiC IMF was proven and the values were compared to conventional UO₂ pellets. XRD studies showed that ceria (PuO₂ surrogate) chemically reacted with the polymer precursor during sintering, forming cerium oxysilicate. Whether PuO₂ will chemically react in a similar manner remains unclear.     

Experimental kinetic study of oxidation of uranium monocarbide powders under controlled oxygen partial pressures below 230 °C

Uranium monocarbide (UC) powders are known to be easily oxidised by gas mixtures containing oxygen. In this study, the oxidation of UC micron powders was followed by isothermal thermogravimetry at temperatures ranging from 100 °C to 230 °C in two different gas mixtures: synthetic air and 97%N₂ + 3%O₂. X-ray diffraction tests conducted on powders after their oxidation showed that small crystallites of UO₂ oxide are formed. Furthermore, an analysis of mass gain showed that the carbon initially linked with the uranium is not oxidised but retained in oxide layers. Additionally, this kinetic study revealed that the rate-limiting step mechanism governing the oxidation of UC powders is a diffusive process that follows the Arrhenius law regarding temperature. Finally, it was discovered that cracks occur in grains once a given fractional conversion has been reached, inducing a major increase in the volume of grains.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.     

Application of a boron doped diamond (BDD) electrode as an anode for the electrolytic reduction of UO2 in Li2O–LiCl–KCl molten salt

A boron doped diamond thin film electrode was employed as an inert anode to replace a platinum electrode in a conventional electrolytic reduction process for UO₂ reduction in Li₂O–LiCl molten salt at 650 °C. The molten salt was changed into Li₂O–LiCl–KCl to decrease the operation temperature to 550 °C at which the boron doped diamond was chemically stable. The potential for oxygen evolution on the boron doped diamond electrode was determined to be approximately 2.2 V vs. a Li–Pb reference electrode whereas that for Li deposition was around ?0.58 V. The density of the anodic current was low compared to that of the cathodic current. Thus the potential of the cathode might not reach the potential for Li deposition if the surface area of the cathode is too wide compared to that of the anode. Therefore, the ratio of the surface areas of the cathode and anode should be precisely controlled. Because the reduction of UO₂ is dependent on the reaction with Li, the deposition of Li is a prerequisite in the reduction process. In a consecutive reduction run, it was proved that the boron doped diamond could be employed as an inert anode.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

Microstructure and properties of tungsten–samarium oxide composite prepared by a novel wet chemical method and spark plasma sintering

W–1 wt% Sm₂O₃ powders doped with highly uniform Sm₂O₃ were successfully synthesized by a novel wet chemical method followed by hydrogen reduction. The powders were consolidated by spark plasma sintering (SPS) at 1800 °C to suppress grain growth during sintering. The FE-SEM and HRTEM analysis, tensile test and thermal conductivity measurements were used to characterize these samples. The grain size, relative density of the bulk samples fabricated by SPS sintering were 4 μm and 97.8%, respectively. The tensile strength values of Sm₂O₃/W samples were higher than those of pure W samples. As the temperature rises from 25 to 800 °C, the thermal conductivity of pure W and W–1 wt% Sm₂O₃ composites decreased with the same trend and the thermal conductivity of both samples was above 160 W/m K at room temperature.     

Is UO2 irradiation resistance due to its unusual high temperature behaviour?

Explaining and predicting the radiation resistance of structural and functional materials is a primary goal for engineering materials able to withstand severe radiation environments. Szenes has developed an empirical criterion based on the thermal behaviour of a compound at high temperature. Though the specific heat at high temperature of most materials obeys the classic Dulong–Petit law, this is not true for uranium dioxide, perhaps the most important ceramic compound in a nuclear power plant. An original analysis of the different contributions to the heat capacity of UO₂ is presented showing that the large increase of UO₂ heat capacity at high temperature (T > 1300 K) is microscopically connected to a high concentration of polarons that are responsible for the departure from the Dulong–Petit law. This is in particular related to the contribution of the uranium sublattice. At the microscopic scale, this thermodynamic anomaly can be related to the thermally activated charge disproportionation of U atoms that is experimentally observed by electrical conductivity measurements. This singular behaviour of the polaron concentration has a direct impact on the uranium sublattice partial molar heat capacity and an indirect effect on the energy interactions between the electronic and ionic structure of the target mediated by these polarons. This could explain, at least partially, the irradiation resistance to amorphisation of UO₂.     

Enhanced arrangement for recuperators in supercritical CO2 Brayton power cycle for energy conversion in fusion reactors

A domestic research program called TECNO_FUS was launched in Spain in 2009 to support technological developments related to a dual coolant breeding blanket concept for fusion reactors. This concept of blanket uses Helium (300 °C/400 °C) to cool part of it and a liquid metal (480 °C/700 °C) to cool the rest; it also includes high temperature (700 °C/800 °C) and medium temperature (566 °C/700 °C) Helium cooling circuits for divertor. This paper proposes a new layout of the classical recompression supercritical CO₂ Brayton cycle which replaces one of the recuperators (the one with the highest temperature) by another which by-passes the low temperature blanket source. This arrangement allows reaching high turbine inlet temperatures (around 600 °C) with medium pressures (around 225 bar) and achieving high cycle efficiencies (close to 46.5%). So, the proposed cycle reveals as a promising design because it integrates all the available thermal sources in a compact layout achieving high efficiencies with the usual parameters prescribed in classical recompression supercritical CO₂ Brayton cycles.     

Comparative study of structural and morphological properties of CuIn3S5 and CuIn7S11 materials

CuIn₃S₅ and CuIn₇S₁₁ powders were prepared by solid-state reaction method using high-purity elemental copper, indium and sulphur. The films prepared from CuIn₃S₅ and CuIn₇S₁₁ powders were grown by thermal evaporation under vacuum (10^?6 Torr) on glass substrates at different substrate temperature Ts varying from room temperature to 200 °C. The powders and thin films were characterized for their structural properties by using X-ray diffraction (XRD) and energy dispersive X-ray (EDX). Both powders were polycrystalline with chalcopyrite and spinel structure, respectively. From the XRD data, we calculated the lattice parameters of the structure for the compounds. For CuIn₃S₅ powder, we also calculated the cation–anion bond lengths. The effect of substrate temperature Ts on the structural properties of the films, such as crystal phase, preferred orientation and crystallinity was investigated. Indeed, X-ray diffraction analysis revealed that the films deposited at a room temperature (30 °C) are amorphous in nature while those deposited on heated were polycrystalline with a preferred orientation along (1 1 2) of the chalcopyrite phase and (3 1 1) of the spinel phase for CuIn₃S₅ and CuIn₇S₁₁ films prepared from powders, respectively. The morphology of the films was determined by atomic force microscopy AFM. The surface roughness and the grain size of the films increase on increasing the substrate temperature.     

Effect of ion-irradiation on the microstructure and microhardness of the W-2Y2O3 composite materials fabricated by sintering and hot forging

A W-2Y₂O₃ material was developed in collaboration with the Plansee Company (Austria). An ingot of the material having approximate dimension of 95 mm × 20 mm was fabricated by mixing the elemental powders followed by pressing, sintering and hot forging. The microstructure of the W-2Y₂O₃ composite was investigated using transmission electron microscopy (TEM). The microhardness was studied using nano-indentation technique. We observed that the W-grains having a mean size of about 1 μm already formed and these grains contain very low density of dislocations. The size of the yttria particles was between 300 nm and 1 μm and the Berkovich hardness was about 4.8 GPa. The specimens were irradiated/implanted with Fe and He ions at JANNuS facility located at Orsay/Saclay, France. The TEM disks kept were irradiated/implanted at 300 and 700 °C using Fe and He ions with an energy of 24 and 2 MeV, respectively. The calculated radiation dose was about 5 dpa produced by Fe ions and total He content is 75 appm at both 300 and 700 °C. From the TEM investigation of irradiated samples, few radiation loops are present on the W grains, whereas on yttria particles, the radiation induced damages appear as voids. Berkovich hardness of the irradiated sample is higher than that of the non-irradiated sample. Results on the microstructure and microhardness of the ion-irradiated W-2Y₂O₃ composites are presented in detail.     

Cerium and neodymium co-precipitation in molten chloride by wet argon sparging

Co-precipitations of cerium (III) and neodymium (III) at 10 wt.% in LiCl–CaCl₂ (30-70 mol%) molten salt at 705 °C have been achieved using an original way of precipitation, wet argon sparging. Several CeCl₃/NdCl₃ ratios have been studied, and the isolated powders were analyzed using different characterization methods including XRD investigations. The lanthanides precipitation yields have been determined around 99.9% using ICP-AES analysis. XRD demonstrated that the precipitates mainly contained mixed oxychloride (Ce_1?xNd_x)OCl and a small amount of the mixed oxide Ce_1?yNd_yO_2?0.5y. Calcination of these precipitates has resulted in the cerium and neodymium mixed oxides. For the precipitation with a Ce/Nd = 50/50 ratio, an hydroxychloride Ln(OH)₂Cl and the oxychloride Ce^IV(Nd_0.7Ce_0.3)^IIIO₃Cl have been identified as unexpected intermediate compounds.     

Manufacturing of self-passivating W-Cr-Si alloys by mechanical alloying and HIP

Self-passivating tungsten-based alloys may provide a major safety advantage in comparison with pure tungsten, which is presently the main candidate material for the plasma-facing protection of future fusion power reactors. WCrSi alloys were manufactured by mechanical alloying (MA) and HIP at 1300 °C and 200 MPa for 1 h. Different MA conditions were investigated to obtain powders with lowest possible amount of contaminants and small and homogeneous particle and crystallite size. Milling in WC vials under Ar without process control agent provided best results. After HIP densities close to 100% were obtained. First oxidation tests on preliminary alloys showed self-passivating behavior with rates comparable to WCrSi thin films at 800 °C but worse performance at 1000 °C. In all cases a Cr₂WO₆ protective layer is formed at the surface.     

Effects of BF2+ implantation on the strain-relaxation of pseudomorphic metastable Ge0.06Si0.94 alloy layers

Metastable pseudomorphic Ge_0.06Si_0.94 alloy layers grown by molecular beam epitaxy (MBE) on Si (1 0 0) substrates were implanted at room temperature by 70 keV BF₂⁺ ions with three different doses of 3 × 10¹³, 1 × 10¹⁴, and 2.5 × 10¹⁴ cm⁻². The implanted samples were subsequently annealed at 800°C and 900°C for 30 min in a vacuum tube furnace. Observed by MeV ⁴He channeling spectrometry, the sample implanted at a dose of 2.5 × 10¹⁴ BF₂⁺ cm⁻² is amorphized from surface to a depth of about 90 nm among all as-implanted samples. Crystalline degradation and strain-relaxation of post-annealed Ge_0.06Si_0.94 samples become pronounced as the dose increases. Only the samples implanted at 3 × 10¹³ cm⁻² do not visibly degrade nor relax during anneal at 800°C . In the leakage current measurements, no serious leakage is found in most of the samples except for one which is annealed at 800°C for 30 min after implantation to a dose of 2.5 × 10¹⁴ cm⁻². It is concluded that such a low dose of 3 × 10¹³ BF₂⁺ cm⁻² can be doped by implantation to conserve intrinsic strain of the pseudomorphic GeSi, while for high dose regime to meet the strain-relaxation, annealing at high temperatures over 900°C is necessary to prevent serious leakages from occuring near relaxed GeSi/Si interfaces.     

Phase diagram of the ZrO2–FeO system

The results on the ZrO₂–FeO system studies in a neutral atmosphere are presented. The refined eutectic point has been found to correspond to a ZrO₂ concentration of 10.3 ± 0.6 mol% at 1332 ± 5 °C. The ultimate solubility of iron oxide in zirconia has been determined in a broad temperature range, taking into account the ZrO₂ polymorphism. A phase diagram of the pseudobinary system in question has been constructed.     

The chemical durability of glass and graphite–glass composite doped with cesium oxide

The role of temperature in determining the chemical stability of a waste form, as well as its leach rate, is very complex. This is because the dissolution kinetics is dependent both on temperature and possibility of different rate-controlling mechanisms that appear at different temperature regions. The chemical durability of Alumina-Borosilicate Glass (ABG) and Glass–Graphite Composite (GGC), bearing Tristructural Isotropic (TRISO) fuel particles impregnated with cesium oxide, were compared using a static leach test. The purpose of this study is to examine the chemical durability of glass–graphite composite to encapsulate coated fuel particles, and as a possible alternative for recycling of irradiated graphite. The test was based on the ASTM C1220-98 methodology, where the leaching condition was set at a temperature varying from 298 K to 363 K for 28 days. The release of cesium from ABG was in the permissible limit and followed the Arrhenius’s law of a surface controlled reaction; its activation energy (E_a) was 65.6 ± 0.5 kJ/mol. Similar values of Ea were obtained for Boron (64.3 ± 0.5) and Silicon (69.6 ± 0.5 kJ/mol) as the main glass network formers. In contrast, the dissolution mechanism of cesium from GGC was a rapid release, with increasing temperature, and the activation energy of Cs (91.0 ± 5 kJ/mol) did not follow any model related to carbon kinetic dissolution in water. Microstructure analysis confirmed the formation of Crystobalite SiO₂ as a gel layer and Cs⁺¹ valence state on the ABG surface.     

Treatment of UO2 pellets used for preparing fuel elements of HTR-10 followed by supercritical fluid extraction

International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO₂ pellets used for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) followed by supercritical fluid extraction was investigated. When UO₂ pellets were dissolved and extracted with tri-n-butyl phosphate (TBP)–HNO₃ complex in supercritical CO₂ (SC-CO₂), the extraction efficiency was less than 7% under experimental conditions. After UO₂ pellets were ground into UO₂ fine powders, the extraction efficiency of the UO₂ fine powders with TBP–HNO₃ complex in SC-CO₂ could reach 92%. After UO₂ pellets broke spontaneously into U₃O₈ powders under O₂ flow and 600 °C, the extraction efficiency of the U₃O₈ powder with TBP–HNO₃ complex in SC-CO₂ could reach more than 98%.     

Fabrication of Li2TiO3 pebbles by a freeze drying process

Li₂TiO₃ pebbles were successfully fabricated by using a freeze drying process. The Li₂TiO₃ slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at ?10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li₂TiO₃ pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Preparation and investigation of aluminized coating and subsequent heat treatment on 9Cr–1Mo Grade 91 steel

Iron aluminide inner coating with alumina top layer is being considered as a potential solution for tritium permeation barrier and mitigating MHD pressure drop for liquid metal blanket concepts in the fusion reactor systems. Hot-dip aluminizing with subsequent heat treatment seems to offer a good possibility to produce aluminized coating with alumina top layer. 9Cr–1Mo Grade 91 steel samples were hot dipped in Al melt containing 2.25 wt% of Si at 750 °C for 3 min. Heat treatment was performed at 650, 750 and 950 °C for 5 h; samples were either air cooled or furnace cooled. Coatings have been evaluated by SEM, EDX, X-ray diffraction, microhardness, scratch adhesion and Raman spectroscopy. The thickness of the layers and phases formed were influenced by the heat treatment adopted. Fe₂Al₅ was the major phase present in the samples heat treated at 650/750 °C, whereas FeAl and α-Fe(Al) primarily made up the outer and inner layers respectively in the samples heat treated at 950 °C. Cooling method deployed affected the hardness. Air cooled samples had comparatively higher hardness than furnace cooled samples. The scratch test showed the adhesion for the samples heat treated at 950 °C was much better as compared to the samples heat treated at 650/750 °C. Raman spectroscopy analysis showed the presence of both α-Al₂O₃ and γ-Al₂O₃ on the surface of the samples heat treated at 950 °C, while Fe₃O₄ was present in the furnace cooled sample only.     

Corrosion behavior of porous chromium carbide in supercritical water

The corrosion behavior of highly porous chromium carbide (Cr₃C₂) prepared by a reactive sintering process was characterized at temperatures ranging from 375 °C to 625 °C in a supercritical water environment with a pressure of 25–30 MPa. The test results show that porous chromium carbide is stable in SCW environments at temperatures under 425 °C, above which disintegration occurred. The porous carbide was also tested under hydrothermal conditions of pressures between 12 MPa and 50 MPa at constant temperatures of 400 °C and 415 °C, respectively. The pressure showed little effect on the stability of chromium carbide in the tests at those temperatures. The mechanism of disintegration of chromium carbide in SCW environments is discussed.