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In the framework of the SELENIUM project, ZrN coated U-Mo fuel kernels were irradiated in the Belgium Reactor 2 of SCK•CEN to a plate average burnup of 48% 235U and a local maximum burnup just below 70% 235U. The microstructure and chemical composition of ZrN coating before and after neutron irradiation have been analysed using scanning transmission electron microscopy (STEM) equipped with energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS), and atom probe tomography (APT). The atomic ratio of N/Zr and fission product distribution determined by three techniques were compared and the combination of three techniques shows the advantages in characterization of chemical information for nuclear materials.

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.

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

Two fuel plates, containing Si and ZrN coated U-Mo fuel particles dispersed in an Al matrix, were irradiated in the BR2 reactor of SCK•CEN to a burn-up of ∼70% 235 U. Five samples were prepared by INL using focused ion beam milling and transported to SCK•CEN for transmission electron microscopy (TEM) investigation. Two samples were taken from the Si coated U-Mo fuel particles at a burn-up of ∼42% and ∼66% 235 U and three samples from the ZrN coated U-Mo at a burn-up of ∼42%, ∼52% and ∼66% 235 U. The evolution of the coating, fuel structure, fission products and the formation of interaction layers are discussed. Both coatings appear to be an effective barrier against fuel matrix interaction and only on the samples having received the highest burn-up and power, the formation of an interaction between Al and U(Mo) can be observed on those locations where breaches in the coatings were formed during plate fabrication.

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.

Grain refinement in U(Mo) has been achieved out-of-pile by irradiating an U(Mo) foil with 80 MeV high-flux 127I ions at 140°C up to a high dose of ~ 2.1 × 1018 ions/cm² or a fission density equivalent of ~ 3.7 × 1021 f/cm³. Several locations in the irradiated area were analysed by scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD) in order to study the dose-dependent evolution of the irradiation induced restructuring. The refined U(Mo) grains formed in proximity of the sample surface, where a strong irradiation induced oxidation was also observed. The spatial concomitance between the irradiation induced grain refinement and sample oxidation points out a possible connection between the two phenomena. The EBSD maps suggest that polygonization was the underlying mechanism driving the grain refinement based on the observation of many low-angle grain boundaries surrounding the newly refined grains. In a later stage, the grain boundaries transformed from low-angle to high-angle under the effect of raising local lattice stresses induced by the ion implantation. The estimated ion irradiation dose triggering the polygonization is approximately one order of magnitude lower than what is observed in-pile. This experiment shows that ion irradiation is capable of simulating the high burnup structure forming in U(Mo) in-pile. Ion irradiation therefore once more proves to be a powerful tool to study the fundamental behaviour of neutron irradiated U(Mo) or other fuel materials.

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175°C (with an estimated error in the range of ±10°C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77±0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results.

The evolution of the high burn-up structure (HBS) in U(Mo) fuel irradiated up to a burn-up of ∼70% 235U or ∼5 × 1021 f/cm3 or ∼120 GWd/tHM is described and compared to the observation made on LWR fuel. Scanning and transmission electron microscopy was performed on several samples having different burn-ups in order to get a better understanding of the mechanisms leading to the high burn-up structure formation. Even though there are some substantial differences between the irradiation of ceramic and U(Mo) alloy fuels (crystal structure, enrichment, irradiation temperature …), it was found that in both fuels recrystallization initiates at the same threshold and progresses in a similar way with increasing fission density. In case of U(Mo), recrystallization leads to accelerated swelling of the fuel which could result in instability of the fuel plate.

The microstructure, growth kinetics, and the high temperature phase equilibria of the U Mn binary system are investigated using differential scanning calorimetry. Alloys with various compositions are prepared in a positive-pressure arc melting furnace in order to examine all possible phase changes within the system. Phase composition and morphology were analyzed using XRD, SEM and EDX. Transformation temperatures and enthalpies have been assessed pertaining to: (i) allotropic phase changes αMn → βMn → γMn →δMn, (ii) two eutectic isotherms UMn2 + βMn → L and U6Mn + UMn2 → L and (iii) melting transitions as a function of composition. The development of the two intermetallic phases U6Mn and UMn2 are observed. A faceted morphology accompanies the formation of UMn2 in a matrix of U6Mn, while grain boundary decomposition is found of UMn2 in αMn. The transformation temperatures are compared to existing phase diagrams and are slightly higher than reported in literature. Experimental observations of the allotropic transformations of αMn to βMn and βMn to γMn are in agreement with a calculated phase diagram. The eutectic composition for UMn2 and αMn is found to be 55 wt%Mn, in agreement with previously reported values, while the enthalpy of fusion associated with this eutectic point is found to be 54 ± 6.5 kJ mol−1. The enthalpies of formation for the two intermetallic phases U6Mn and UMn2 are 71 ± 8.5 kJ mol−1 and 42 ± 5 kJ mol−1 respectively.

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.

Irradiation responses of U–Mo/Al dispersion fuel have been investigated by irradiation with 84 MeV Xe26+ ions. Dispersion fuels fabricated with uncoated and ZrN-coated fuel particles were irradiated to various doses at ∼350 °C. The highest dose achieved was 2.9 × 1017 ions/cm2 (∼1200 displacement per atom (dpa)). Following the irradiation, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) experiments were carried out to characterize the microstructures of the irradiated samples. The post irradiation examinations (PIE) revealed that: (1) crystalline interdiffusion product (UMo)Alx developed at locations where no coating or compromised coating layer is present; (2) intact ZrN coating layers effectively blocked the interdiffusion between U–Mo and Al; (3) SEM-observable Xe bubbles distributed along grain/cell boundaries in U–Mo; and (4) gas bubble interlinkage was observed at a dose of 2.9 × 1017 ions/cm2.

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.

This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

A new type of high density fuel is needed for the conversion of research and test reactors from high to lower enriched uranium. The most promising one is a dispersion of atomized uranium-molybdenum (U–Mo) particles in an Al matrix. However, during in-pile irradiation the growth of an interaction layer between the U–Mo and the Al matrix strongly limits the fuel’s performance. To improve the in-pile behaviour, the U–Mo particles can be coated with protective layers. The SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) fuel development project consists of the production, irradiation and post-irradiation examination of 2 flat, full-size dispersion fuel plates containing respectively Si and ZrN coated U–Mo atomized powder dispersed in a pure Al matrix. In this paper X-ray diffraction analyses of the Si and ZrN layers after deposition, fuel plate manufacturing and thermal annealing are reported. It was found for the U–Mo particles coated with ZrN (thickness 1 μm), that the layer is crystalline, and exhibits lower density than the theoretical one. Fuel plate manufacturing does not strongly influence these crystallographic features. For the U–Mo particles coated with Si (thickness 0.6 μm), the measurements of the as received material suggest an amorphous state of the deposited layer. Fuel plate manufacturing strongly modifies its composition: Si reacts with the U–Mo particles and the Al matrix to grow U(Al, Si) 3 and U 3 Si 5 phases. Finally both coatings have shown excellent performances under thermal treatment by limiting drastically the U–Mo/Al interdiffusion.

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.

Several irradiation experiments have confirmed the positive effect of adding Si to the matrix of an U(Mo) dispersion fuel plate on its in-pile irradiation behavior. E-FUTURE, the first experiment of the LEONIDAS program, was performed to select an optimum Si concentration and fuel plate heat treatment parameters for further qualification. It consisted of the irradiation of 4 distinct (regarding Si content and heat treatments), full size flat fuel plates in the BR2 reactor under bounding conditions (470 W/cm 2 peak BOL power, ∼70% peak burn-up). After the irradiation, the E-FUTURE plates were examined non-destructively and found to have pillowed in the highest burn-up positions. The destructive post-irradiation examination confirmed that the fuel evolves in a stable way up to a burn-up of 60% 235 U. Even in the deformed area (pillow) the U(Mo) fuel itself shows stable behavior and remaining matrix material was present. From the calculation of the volume fractions, the positive effect of a higher Si amount added to the matrix and the higher annealing temperature can be derived.

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.

Excessive U-Al interdiffusion hampers the development of UMo/Al based disperse fuel since several years. As a remedy, addition of diffusion limiting elements to the UMo (Zr) or the Al matrix (Si) has been suggested. First test irradiations showed promising results. However, not the full amount of third element addition to the UMo or the Al contributes to the prevention of interdiffusion layer formation. Therefore, surface engineering of UMo particles, i.e. coating of UMo powder with diffusion limiting elements using DC magnetron sputtering, has been suggested. Thereby the diffusion blockers are applied directly where they are needed: at the interface between the UMo and the Al. For this study, samples from full size plates produced with UMo powder coated with Si (300 nm and 600 nm) and ZrN (1000nm) have been examined before and after irradiation with Iodine at 80 MeV. In case of Si coated UMo particles, a dense Si rich layer developed around the UMo particles during plate fabrication. However, the ZrN layer frequently revealed cracks but no interaction with the Al matrix or the UMo occurred. During heavy ion irradiation, no UMo/Al interdiffusion occurred at spots that are protected by a sufficiently thick Si rich layer or a non-cracked ZrN layer. In contrast, a conventional UMo/Al interdiffusion layer (IDL) occurred at spots where the UMo has not been protected by a Si rich layer or the ZrN layer was broken.

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 fusion devices, in the region next to high temperature plasma, material could be eroded from plasma-facing materials in one location and is transported to other, sometimes remote, locations throughout the device. These plasma-created materials are so-called mixed materials. Previous studies on JET samples have shown that compositional changes in the depth profiles of the mixed-material surfaces can be correlated to the operational history of the machine. SEM and detailed EDX analysis of CFC Be and T contaminated samples coming from the JET divertor (shutdown 2007) have been carried out in the SCKCEN hot cell facilities. The main aim of these measurements is to have a clear image of the erosion/deposition pattern in those samples for each element. The processes for material migration and the morphology of the deposits are discussed. XPS/AES measurements will also be carried out to understand the possible compound processes, mainly focussed on Be 2 C formations.

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.

In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235 U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235 U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900–950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

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.

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 microstructural evolution of atomised U–7 wt%Mo alloy fuel under irradiation was investigated by transmission electron microscopy on material from the experimental fuel plates used in the FUTURE irradiation. The interaction layer that forms between the U(Mo) particles and the Al matrix is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, causing the fuel plate to pillow and finally fail. The present analysis confirms the assumption that the U(Mo)–Al interaction layer is completely amorphous after irradiation. The Al matrix and the individual U(Mo) particles, with their cellular substructure, have retained their crystallinity. It was furthermore observed that the fission gas generated in the U(Mo) particles has formed a bubble superlattice, which is coherent with the U(Mo) lattice. Bubbles of roughly 1–2 nm size have formed a 3-dimensional lattice with a lattice spacing of 6–7 nm.

Three full size AlFeNi cladded U3Si2 fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCKCEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCKCEN. The PIE shows that the fuel plates withstood the irradiation success- fully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U3Si2 fuel and the Al matrix, has been quantified as U(Al,Si)4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U3Si2 are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U3Si2 the bubbles are larger but there are fewer. 2008 Elsevier B.V. All rights reserved.

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.

PH13-8Mo bolts, which are considered for use in the ITER reactor, were irradiated up to doses of 0.5, 1 and 2 dpa. The microstructure was investigated with transmission electron microscopy and its evolution is discussed with reference to the mechanical properties. PH13-8Mo is a precipitation hardened martensitic steel, but a large amount of austenite has been observed as well. The precipitation hardening results from the formation of small coherent NiAl precipitates in the martensite phase. Their size, size distribution and density are found to be unaffected by neutron irradiation. The dislocations in the martensite phase are mainly a/2h 111 i type screw dislocations, whereas in the austenite phase mainly a/2h 110 i type screw dislocations are present. The line dislocation structure did not change during irradiation, but small irradiation induced defects were observed. Using the Orowan model, it is argued that the latter are responsible for the irradiation hardening.

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).

In order to gain insight on the performance limits of U 3 Si 2 fuel with Al cladding, fuel plates with a fissile material density of 5.1 and 6.1 g U/cm 3 were irradiated in the BR2 reactor of SCK • CEN in Mol. The plates were intended to be subjected to severe conditions leading to a cladding surface temperature of 180–200 °C and fuel temperatures of 220–240 °C. The irradiation program was stopped after the second cycle based on the visual inspection and wet sipping tests of the elements, correspondingly showing degradations on the outer Al surfaces of the U 3 Si 2 plates and the release of fission products. The maximum fuel burn-up was 29% and 25% 235 U, respectively. In a PIE program the microstructural causes for this degradation are analysed. It is found that the failure of the plates is entirely related to the corrosion of the Al cladding, which has caused temperatures to rise well beyond the calculated values. In all stages, the fuel grains have retained their integrity and, apart from the formation of an interaction phase with the Al matrix, they do not demonstrate deleterious changes in their physical properties.

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.

Spent fuel remnants, cut from BWR fuel rods were retrieved for microscopic investigations after 25 years of storage at ambient temperature. The storage atmosphere of the samples studied here was dry air for 10 years and a moist environment for the remaining 15 years. The comparison with earlier experiments conducted on samples that were stored in dry air for the full period demonstrates that moisture accelerates the oxidation process also at the low temperatures considered here. The composition of the atmosphere in contact with the fuel has been analysed by mass spectrometry and the microstructure of the samples has been investigated by optical and scanning electron microscopy. The microscopic observations evidence that grain boundaries are affected and different stages in grain boundary alteration could be evidenced. Oxidation first induces a weakened intergranular bond followed by increased sensitisation to chemical attack and finally decohesion and onset of bulk oxidation.

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.

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 .