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Image analysis was performed on SEM micrographs of recently irradiated UMo dispersion fuel plates. Detailed information on the fission gas inter-granular bubbles accompanying recrystallization, including average diameter, density and size distribution were extracted and compared with previous results on UMo and UO2. A pore density drop was notice at high fission densities and attributed mainly to a pore coarsening dominated by irradiation induced phenomena. Based on the image analysis data and theoretical considerations, a model was developed to estimate the pressure inside the pores as a function of fission density, temperature and pore radius. The developed pressure can give indications of the mechanical stability of the fuel towards the progressive building-up of fission gases. Finally, the proposed methodology was applied to the nanobubble lattice decorating the fuel grains at low fission densities in order to infer the physical state of the contained fission gases. The estimated values suggest the presence of solid xenon precipitates.

The control of the interaction between the U(Mo) fuel phase and the Al matrix is one of the challenges of dispersion fuel plate development for research reactors. Given the specific properties of this interaction layer, larger amounts of it in the meat could lead to a reduction of the plate mechanical integrity and thermal conductivity, eventually leading to pillowing. The SELENIUM project showed that by depositing a ZrN coating on the surface of the U(Mo) fuel particles, the amount of formed U(Mo)-Al interaction layer is limited but still present. Microstructural analysis performed on the as fabricated coating and fresh fuel plates containing ZrN coated U(Mo) dispersed in an Al matrix, revealed that the coating gets damaged during plate production. The post irradiation examinations (PIE) of the ZrN coated U(Mo) fuel plates, from the SELENIUM and SEMPER FIDELIS experiments, show how the U(Mo)-Al interaction layer is formed - only at those locations where the coating is missing or damaged - and the evolution of coating microstructure during irradiation. As a remedy, to further reduce the amount of interaction layer formed, the use of an Al-Si matrix was proposed based on the higher affinity of Si for U compared to the affinity of Al for U. PIE of a fuel plate consisting of ZrN coated U(Mo) dispersed in an Al-Si matrix irradiated in the SEMPER FIDELIS experiment, clearly demonstrates the benefit of adding Si to the matrix.

Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK·CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm2) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm2, throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%235U compared to 70%235U for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

Aluminum cladding oxidation of research-reactor fuel elements at high power conditions has a disadvantageous effect on fuel performance due to the lower thermal conductivity of the oxide. The oxide growth prediction models available in the literature were mostly developed for low power conditions. To examine the applicability of the models to high power and high temperature test conditions, the models were studied by coupling with the most frequently employed heat transfer coefficient (HTC) correlations including the Dittus-Boelter correlation, the Colburn correlation, the Sieder-Tate correlation, and KAERI-developed correlation. The Griess model over-predicted the oxide growth while the KAERI-Griess model under-predicted the oxide growth for high power tests. The Kim model, coupled with the Colburn correlation, gave most consistent results with the measured data from two BR2 experiments. However, the Kim model was found to be inapplicable to the EUHFRR conditions at the peak power locations if it was coupled with the Dittus-Boelter correlation. A revision of the prediction models to more closely agree with the measured data was recommended. AG3NE and AlFeNi cladding types were tested in the E-FUTURE experiment, and a noticeable (although small) reduction in oxide thickness on the AlFeNi cladding was observed. However, this difference was believed to be only a secondary effect considering other uncertainties in model predictions, so no attempt was made to model the alloying effect.

Metallic alloy U-Zr nuclear fuels remain a candidate for future US fast reactors. However, with more than 30 years of investigation into the system, U-Zr fuel remains unqualified and lacks predictive modeling capabilities. Phenomenon identification and ranking tables (PIRT) were developed for U-Zr fuels to prioritize the study of microstructural phenomena related to fuel performance and to aid predictive models. Two areas of emphasis, fuel swelling and constituent redistribution, were identified with PIRT and were evaluated in detail. The analysis identified influencing parameters that should be considered during future studies including: external mechanical constraint, temperature, composition, crystallographic texture, temperature gradient, fabrication constraints, fission density, fission rate, initial porosity, and pre-irradiation grain size. The PIRT results support the need for a comprehensive study that focus on the decoupling of fission rate, irradiation temperature, and alloy composition to independently investigate the microstructural evolution in the multiple phases typical of advanced reactor conditions.

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235 U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK⋅CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm2 peak Beginning Of Life (BOL) power, with a ∼70% 235U peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 × 1021 f/cm3, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article.

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235 U. The root cause of the failures is clearly related directly to the formation of the U(Mo)–Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm 2 until a maximum local burn-up of approximately 70% 235 U (∼50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

Several irradiation experiments of U(Mo) dispersion fuel performed with aluminum as matrix resulted in unacceptable swelling of the fuel plate due to the formation of an interaction layer between Al and U(Mo). It was found that an improvement in fuel behavior can be achieved by adding Si to the Al matrix and creating a Si rich preformed layer which delays the formation of the interaction layer. Such Al–Si matrices can be formed either by mixing silicon powder with aluminum or using an AlSi alloy. AlSi alloy powders have very different mechanical properties which complicate fuel plate fabrication. Aging experiments on AlSi alloys reveal that giving the alloy the correct heat treatment results in a homogenous dispersion of fine Si precipitates in a soft and strain free Al matrix. The diffusion of such small precipitates towards the U(Mo) particles will be more effective than the transportation of Si from the larger Si particles used in a mixture matrix. Out of pile experiments are performed to show the difference between using a mixture or an alloy for the interaction with U(Mo). It was found that the U(Mo) particles dispersed in an AlSi alloy matrix have a more uniform Si rich preformed layer after heat treatment.

In the framework of the elimination of High-Enriched Uranium (HEU) from the civil circuit, the search for an appropriate fuel to replace the high-enriched research reactor fuel in those reactors that currently still require it for their operation has led to the development of a U–7 wt.%Mo alloy based dispersion fuel with an Al–Si matrix. The European LEONIDAS program, joining SCK•CEN, ILL, CEA and AREVA-CERCA, is aimed at the qualification of such a fuel for the use in high power conditions. The first experiment of the program, designated E-FUTURE, was performed to select the appropriate matrix Si concentration and fuel plate post-production heat treatment parameters for further qualification. It consisted of the irradiation of four distinct (4% and 6% Si, 3 different heat treatments) full size, flat fuel plates in the BR2 reactor. The irradiation conditions were relatively severe: 470 W/cm 2 peak BOL power, with a ∼70% 235 U peak burnup.

This paper focuses on the modelling of fission gas release in mixed oxide fuel. In a first part, the irradiation experiment, conducted by the Japanese Nuclear Energy Safety organization on high enriched mixed oxide fuel, is outlined. In a second part, the approach for fission gas release modelling, as implemented in the fuel performance code MACROS, is explained and a comparison between calculated and experimental results is made. The code MACROS is conceived to provide not only integral (rod average) results on fission product and fission gas retention and release, but also to calculate local concentrations (radial profiles). In this way, it is possible to compare results from post-irradiation examinations with calculated profiles.

Chromia doped UO2 fuel was characterised by electron microprobe analysis (EPMA) and X-ray powder diffraction (XRD). The macroscopic distribution of chromium across the fuel pellet (homogeneity, surface evaporation effect) and its microscopic state (extent of atomic scale dissolution in the UO2 matrix versus appearance as precipitates) was investigated. Furthermore, the lattice parameter obtained by X-ray powder diffraction was calculated by the unit cell refinement method and compared to empirical interatomic potential calculations (energy minimisation techniques). Finally, the theoretical density of the chromia doped fuel was calculated.

In the framework of the IRIS-TUM irradiation program, several full size, flat dispersion fuel plates containing ground U(Mo) fuel kernels in an aluminum matrix, with and without addition of silicon (2.1 wt.%), have been irradiated in the OSIRIS reactor. The highest irradiated fuel plate (with an Al–Si matrix) reached a local maximum burnup of 88.3% 235 U LEU-equivalent and showed a maximum thickness increase of 323 μm (66%) but remained intact. This paper reports the post irradiation examination results obtained on four IRIS-TUM plates. The evolution of the fission gas behavior in this fuel type from homogeneously dispersed nanobubbles to the eventual formation of large but apparently stable fission gas bubbles at the interface of the interaction layer and the fuel kernel is illustrated. It is also shown that the observed moderate, but positive effect of Si as inhibitor for the U(Mo)–Al interaction is related to the dispersion of this element in the interaction layer, although its concentration is very inhomogeneous and appears to be too low to fully inhibit interaction layer growth.

Ruthenium films were grown by plasma enhanced atomic layer deposition (ALD) on Si(100) and ALD TiN. X-ray diffraction (XRD) showed that the as-deposited films on Si(100) were polycrystalline, on TiN they were (002) oriented. After annealing at 800^oC for 60s, all Ru films were strongly (002) textured and very smooth. Electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) demonstrated that the lateral grain size of the annealed films was several 100nm, which was large compared to the 10nm thickness of the films. No ruthenium silicide was formed by annealing the ALD Ru films on Si(100). Comparison with sputter deposited films learned that this occurred because the ammonia plasma created a SiO"xN"y reaction barrier layer prior to film growth.

The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma wall-interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In JET facility both beryllium and carbon are used together as plasma facing materials, and the resultant surfaces show considerable mixing between the two elements. In this work a first XPS characterisation has been performed on the surface of several PFC (plasma facing components) samples obtained from various JET tiles from the Mk II GB divertor in JET. The substrates of the tiles are CFC, and the deposited films have various D/C ratios depending on the position in the tokamak. Ion beam analysis shows that the D/C ratio is ∼1 in the shadowed areas and Be/C ∼1–2 in the plasma-exposed regions of the inner divertor. These values could give rise to the formation of beryllium carbides, and indications of a carbide component are found in the XPS spectra. The processes for material migration and the morphology of the deposits are discussed.

Valence states of uranium in three ternary uranates KUO 3 , NaUO 3 and Ba 2 U 2 O 7 were studied using X-ray photoelectron spectroscopy (XPS). The analysis of the U4f photoelectron peak positions, peak widths at half maximum and satellite positions of U4f peaks, shows that the uranium in these compounds is present in a single valence state, namely U 5 + . Since the surface of these compounds is extremely susceptible to oxidation, the XPS spectra were measured after a carefully controlled etching of the surface using an Ar + ion beam so that the surface oxidation does not affect the XPS studies.

As swipe samples from enrichment activities typically contain uranium particles with a detectable amount of fluorine, the question was raised whether the analysis of fluorine in particles could complement the information on the uranium isotope ratios. For this, uranium oxyfluoride particles were prepared from the controlled hydrolysis of uranium hexafluoride (UF 6 ). The relative amount of fluorine was characterized by scanning electron microscopy combined with energy-dispersive X-ray spectrometry (SEM-EDX), as well as ion-microprobe secondary ion mass spectrometry (IM-SIMS). Of particular interest was the assessment of the reduction of the amount of fluorine over time, and after exposure to UV-light and high temperatures. Micro-Raman spectrometry (MRS) was applied to look for differences in molecular structure between these various sample types. Both SEM-EDX and IM-SIMS showed a general reduction of the fluorine-to-uranium ratio after 1–2 years of storage. The exposure to UV-light and high temperatures was found to have accelerated the loss of fluorine. A distinct peak at 865 cm − 1 Raman shift was detected for the majority of particles analyzed by MRS. For the particles that were heat-treated, the Raman spectra were similar to the spectrum of U 3 O 8 . Although often large variations were observed between particles from the same sample, the three particle measurement techniques (IM-SIMS, SEM-EDX and MRS) showed some consistent trends. They therefore appear promising in terms of the ability to place bounds on particle age, as well as shedding light on the complex processes involved in UO 2 F 2 particle ageing.

TiN was grown by atomic layer deposition (ALD) from tetrakis(dimethylamino)titanium (TDMAT). Both thermal and plasma enhanced processes were studied, with N"2 and NH"3 as reactive gases. Using an optimized thermal ammonia based process, a growth rate of 0.06nm/cycle and a resistivity of 53x10^[email protected]@Wcm were achieved. With an optimized plasma enhanced NH"3 process, a growth rate of 0.08nm/cycle and a resistivity of 180 @[email protected] cm could be obtained. X-ray photo electron spectroscopy (XPS) showed that the difference in resistivity correlates with the purity of the deposited films. The high resistivity of thermal ALD films is caused by oxygen (37%) and carbon (9%) contamination. For the film deposited with optimized plasma conditions, impurity levels below 6% could be achieved. The copper diffusion barrier properties of the TiN films were determined by in-situ X-ray diffraction (XRD) and were found to be as good as or better than those of films deposited with physical vapor deposition (PVD).

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si)3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si)3 layer.

The Material Testing Reactor BR2 of SCK·CEN has a beryllium matrix which has been replaced in 1995. Pieces of this matrix, which were irradiated over a period of 15 yr at about ±50 °C up to a fast neutron fluence ( E > 1 MeV) of 4.67 × 10 22 n/cm 2 , have been retrieved for creep and microstructure studies. An amount of helium resulting from transmutation has accumulated up to approximately 2.2 at.% which is comparable to the expected He content of a fusion reactor blanket at its end of life. Post-irradiation anneals have been performed at 500, 750 and 900 °C, with an annealing time ranging between 50 hours and three months. Microstructural analyses show a clear evolution in bubble growth and He diffusion towards the grain boundaries where the bubbles coalesce. Complete release of 4 He is observed at temperatures >750 °C, resulting in a stagnation of the beryllium swelling. The microstructural changes are mainly temperature driven. Only at high annealing temperature, an influence of the annealing time is observed.

Two low-enriched uranium fuel plates consisting of U–7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm 2 while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235 U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al 3 and (U,Mo)Al 4 . Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat a l’energie atomique (CEA).

The solubility of uranium dioxide (UO2) was measured in real and synthetic Boom Clay waters with varying concentrations of humic acids and carbonate under reducing conditions at 20 C. Uranium concentrations in function of time suggest the reduction of U(VI) to U(IV) by the humic acids which is occurring faster in real clay water than in synthetic clay waters. Humic acids induce also a competition to complex U(VI) in carbonate-containing solution, but they are not able to control the uranium concentration at high bicarbonate concentration (0.02 mol dm � 3 ). Nevertheless they may play a role at low carbonate concentration. In our experimental conditions, the geochemical calculations indicate that two uranium secondary phases (U4O9 and UO2ðcÞ) are susceptible to control the uranium concentration in solution. These calculations are in good agreement with results of the X-ray photoelectron spectroscopy. At the end of tests, uranium concentrations reach steady-state values between 3 � 10 � 8 and 5 � 10 � 8 mol dm � 3 in the bicarbonate-rich solutions. Although these concentrations are considered as conservative, they are 10–100 times higher than in natural Boom Clay. The consequence is that spent fuel could slowly dissolve in the interstitial clay water undersaturated with respect to UO2/UO2þx of the fuel. 2003 Elsevier B.V. All rights reserved.

Chromium sesquioxide doped uranium dioxide systems with a variety of initially added Cr2O3 amounts were prepared under different sinteringconditions. The solubility limit of Cr for each sinteringcondition is derived from the measured content of the dopant dissolved in the UO2 matrix usingelectron probe microanalysis (EPMA). It is found that the solubility of chromium in a uranium dioxide system for these series sintered at 1600, 1660 and 1760 C is limited to respectively 0:065 � 0:002, 0:086 � 0:003 and 0:102 � 0:004 wt% Cr. The lattice parameters of the different Cr2O3 doped fuels have been examined with X-ray diffraction (XRD). The XRD peaks of the samples sintered at 1760 Ca s well as a UO2 reference without Cr, prepared under the same conditions, were measured and a value for the lattice parameter a for each sample was obtained usingthe unit cell refinement method. A slig ht contraction of the lattice parameter is observed with increasingdopant content. 2003 Elsevier Science B.V. All rights reserved.

The use of sputter deposited thin layers of UO 2 as a model system for the investigation of fuel-fission product interactions is presented. The representativity of the layers for the bulk system will be validated and it will be shown, both on theoretical and experimental grounds, that layers of stoichiometric UO 2 can be produced by this method. A comparison will be made between X-ray photoelectron spectroscopic (XPS) results on bulk UO 2 and on the deposited layers. The films deposited can easily be doped with other elements, such as fission products, by codepositing these elements with the UO 2 . This codeposition technique has subsequently been used to produce layers of UO 2 containing cesium. It will be demonstrated that the codeposition with cesium produces uranium in higher valence states (up to U VI ), while without cesium, no higher uranium valencies can be obtained.

Silicon or germanium-based transistors are nowadays used in direct contact with silicide or germanide crystalline alloys for semiconductor device applications. Since these compounds are formed at elevated temperatures, accurate knowledge of the thermal expansion of both substrate and the contact is important to address temperature depending effects such as thermal stress. Here we report the linear coefficients of thermal expansion of Ni-, Pd- and Pt-based mono-germanides, mono-silicides and di-metal-silicides as determined by powder-based x-ray diffraction between 300 and 1225 K. The investigated mono-metallic compounds, all sharing the MnP crystal structure, as well as Pd2Si and Pt2Si exhibit anisotropic expansion. By consequence, this anisotropic behaviour should be taken into account for evaluating the crystal unit's cell at elevated temperatures.

We investigated the local atomic and electronic structures of two related systematic sets of ternary uranium oxides, NaUO3–KUO3–RbUO3 and BaUO3–Ba2U2O7–BaUO4, by measuring the X-ray absorption near edge structure (XANES). The results are compared with calculations based on a self-consistent real space full multiple scattering analysis. We found a very good agreement between measured and calculated spectra, which indicates that the uranium ions are in a pure U5+ oxidation state in these compounds. The low energy shoulder observed in the U L3 edge XANES is an intrinsic feature of the uranium unoccupied 6d electronic states of the U5+ ions within the studied materials. Specific double shoulder features in the higher energy range of the U L3 edge XANES can be interpreted as indicative of the pure cubic perovskite structure.

Several pure cesium uranate phases were prepared (Cs 2 U 2 O 7 , Cs 2 U 4 O 13 , Cs 2 U 4 O 12 , Cs 4 U 5 O 17 ) and measured with X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS). Detailed XPS scans of the main photoelectron peaks (U 4f, Cs 3d, O 1s and valence band) were made. Peak positions, peak widths and satellite positions were measured and compared between the uranates. A systematic shift in the position of the U 4f lines was found to be correlated with the Cs/U ratio. The mixed valency behaviour of uranium was observed and studied. Interpretation of the satellite structure of the U 4f peaks and the deconvolution of these peaks have been used to quantify the chemical states of uranium present in the mixed valency compound Cs 2 U 4 O 12 .

Vanadium pentoxide was deposited by atomic layer deposition (ALD) from vanadyl-tri-isopropoxide (VTIP). Water or oxygen was used as a reactive gas in thermal and plasma-enhanced (PE) processes. For PE ALD, there was a wide ALD temperature window from 50 to 200°C. Above 200°C, VTIP decomposed thermally, resulting in the chemical vapor deposition (CVD) of vanadium pentoxide. The PE ALD reactions saturated much faster than during thermal ALD, leading to a growth rate of approximately 0.7 A/cycle during PE ALD using H 2 O or O 2 . Optical emission spectroscopy showed combustion-like reactions during the plasma step. X-ray diffraction was performed to determine the crystallinity of the films after deposition and after postannealing under He or O 2 atmosphere. Films grown with CVD at 300°C and PE O 2 ALD at 150°C were (001)-oriented V 2 O 5 as deposited, while thermal and PE H 2 O ALD films grown at 150°C were amorphous as deposited. The crystallinity of the PE O 2 ALD could be correlated to its high purity, while the other films had significant carbon contamination, as shown by X-ray photoelectron spectroscopy. Annealing under He led to oxygen-deficient films, while all samples eventually crystallized into V 2 O 5 under O 2 .