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Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like ²⁴¹Am is therefore an option for the reduction of radiotoxicity and heat production of waste packages to be stored in a repository. The MARINE irradiation experiment is the lat...
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... These models are actually developed for U-Pu MOX, and are still selected in light of the low (homogeneous, within [0 ÷ 5] wt.%) Am concentration in the as-fabricated fuel currently considered for MYR-RHA applications. This results in a limited impact on the mechanical properties compared to U-Pu MOX (Sobolev et al., 2003;Kato et al., 2011;Prieur et al., 2015), hence the models for MOX elastic moduli and strain due to thermal expansion can be deemed suitable also for Am-MOX (while a dedicated modelling should be necessary to target oxide fuels bearing higher amounts of minor actinides, for e.g., blanket fuel and transmutation purposes (D'Agata et al., 2017)). This is in line with the state-of-the-art approach for the modelling available in fuel performance codes (Van Uffelen and Suzuki, 2012;Van Uffelen et al., 2019), and the same strategy is adopted for all the other, several fuel properties not mentioned before (e.g., specific heat, creep). ...
Minor actinides are the main contributors to medium- and long-term radiotoxicity and heat production in spent nuclear fuels. Research efforts are currently ongoing to explore different options to dispose of such radionuclides, e.g., their burning in fast reactors within mixed-oxide fuels. The MYRRHA sub-critical reactor is one of the future facilities with envisaged burning and transmutation capabilities. This work assesses the thermal–mechanical performance of a homogeneous Am-bearing fuel pin both in the In-Pile test Section position of the MYRRHA “Revision 1.8” core and under driver irradiation. The normal operating conditions of MYRRHA are considered, with a focus on the safety design limits and involving sensitivity analyses to evaluate the impact of increasing americium contents (in the range 0–5 wt.%) on safety-relevant simulation outcomes. The simulations are performed with the TRANSURANUS fuel performance code (version v1m4j22) coupled with the SCIANTIX physics-based module for inert gas behaviour, and rely on a dedicated surrogate model for the helium source term during MYRRHA irradiation accounting for the relevant contribution of the fuel americium enrichment, besides advanced models for the properties and behaviour of the specific pin materials. The analyses reveal the suitability and safety under irradiation of MOX fuels with low Am enrichments according to the current MYRRHA design.
... Within the SPHERE irradiation experiment a mixedoxide particle fuel (composition U 0.76 Pu 0.20 Am 0.03 O 2-x ) was developed by infiltration of porous beads prepared from external and internal gelation with americium solution [25] . In the later MARINE irradiation experiment the fabrication of mixed-oxide fuels with Am contents up to 14% metal fraction similarly employed infiltration, but the precursors were made using external gelation only [26] . Recently, the application of infiltration on internally gelled UO 3 microspheres was also reported for manganese and chromium doping (up to 20 0 0 ppm, dopant to uranium weight concentration) [27] . ...
Process conditions for the fabrication of porous uranium oxide microspheres prepared via internal gelation were assessed. To improve conditions for the application of infiltration, microstructural parameters such as density, porosity and specific surface area were assessed. Specifically, the effect of calcination temperature and the use of pore-formers was studied. Accessible porosity levels around 20% were obtained after calcination at 773 K or 823 K, without the use of a pore-former. As a novel application, starch was used as a low-temperature, burnable pore-former, and its effect was compared to that of graphite. Accessible porosity levels increased to 34% after calcination due to the use of starch, whereas the application of graphite was discarded because it requires too elevated calcination temperatures. A subset of porous uranium oxide microspheres was infiltrated with neodymium nitrate solution as a surrogate for americium nitrate. Very good agreement between targeted and actual Nd content was observed after sintering of the microspheres, and a maximum concentration of y = 25 mol% (U1-yNdyO2-x) could be reached.
... The isotope decays primarily via alpha decay to Np-237. In a final repository Am241 is one of the most important drivers for the medium-term heat load, and therefore the required gallery space, and Np-237 (t 1/2 =2.14•10 6 y) is one of the significant isotopes driving longterm radiotoxicity [40,41]. The main radiation of Am-241 are alpha particles with an average energy of 5.47 MeV. ...
... The capsules were welded using established Tungsten Inert Gas welding equipment, which is also used in the frame of qualified welding of fuel rodlets for irradiation experi ments [41]. Non-destructive as well as destructive weld examinations were performed and showed that good welding results were achieved; indicating that future welding quality criteria can be met ( Figure 13). ...
... Zirconium (Zr) alloys are extensive applied for cladding materials and spacer grid for pressurized water reactors (PWR) because of low neutron capture cross section and good mechanical strength [1][2]. In high temperature and pressure water condition of reactor, spacer grids directly contact and fasten with Zr alloys cladding tube [3][4]. While in service, the cladding tubes are frequently subjected to axial and lateral vibration generated by high-speed flow of coolant in reactor, called as Flow-induced vibration (FIV). ...
This paper details the first dedicated production of homogeneous nanocrystalline particles of mixed actinide oxide solid solutions containing americium. The target compositions were U0.75Pu0.20Am0.05O2, U0.90Am0.10O2 and U0.80Am0.20O2. After successful hydrothermal synthesis and chemical characterisation, the nanocrystals were sintered and their structure and behaviour under self-irradiation were studied by powder XRD. Cationic charge distribution of the as-prepared nanocrystalline and sintered U0.80Am0.20O2 materials was investigated applying U M4 and Am M5 edge high energy resolution XANES (HR-XANES). Typical oxidation states detected for the cations are U(iv)/U(v) and Am(iii)/Am(iv). The measured crystallographic swelling was systematically smaller for the as-synthesised nanoparticles than the sintered products. For sintered pellets, the maximal volumetric swelling was about 0.8% at saturation, in line with literature data for PuO2, AmO2, (U,Pu)O2 or (U,Am)O2.
Uranium-bearing microspheres below 50µm with a narrow size distribution allows for a wider variety of fuel forms. To accommodate the smaller size, gel microspheres with a composition of UO3∙nH2O∙mNH3 were synthesized using microfluidics and subsequently converted to U3O8 and UO2. To accommodate the slower flow rates required by microfluidics, a more stable broth was established. The gelation studies resulted in a broth that was stable for more than two days at 0°C and for close to 3 hours at room temperature while still gelling within 25 seconds. Synthesis of gel microspheres with a narrow size distribution lasted for 5 hours and produced ∼0.5 g of air-dried material. The gelled microspheres were converted to U3O8 and UO2 and with sizes of 50 and 40 µm in diameter, respectively.
Recycling and burning minor actinides (MA, e.g., americium, neptunium) in mixed-oxide (MOX) nuclear fuel is a strategic option for fast reactor concepts of Generation IV, especially considering the current interest in the ultimate radioactive waste management and sustainability improvement by better use of natural resources. Among the fuel properties, thermal conductivity and melting temperature are pivotal since they determine, respectively, the fuel temperature profile and the fundamental safety limit on the margin to fuel melting, hence impacting on the overall fuel performance under irradiation and allowing the safe irradiation of the fuel pin. Nevertheless, the available literature about Am- or Np-containing MOX is currently scarce, both regarding experimental data and models. Moreover, state-of-the-art fuel performance codes (FPCs, e.g., TRANSURANUS) do not account for the effects of minor actinides on MOX fuel properties. This work presents original correlations for thermal conductivity and melting temperature of minor actinide-MOX fuels, i.e., (U, Pu, Am, Np)O2-x, derived based on the available literature and accessible data, which are herein extensively reviewed. The assessment of the novel correlations is first performed in a statistical way, evaluating the regressor p-values which indicate their significance with respect to the available experimental dataset used for the fitting procedure. Additionally, the novel correlations for MA-MOX are assessed against both measured and calculated data (from Molecular Dynamics simulations), yielding an accuracy in line with the already existing correlations and with the state-of-the-art experimental uncertainties. Finally, the potential integral impact of a homogeneous minor actinide content in the fuel is illustrated on the basis of a fuel pin fast-ramped up to fuel melting during the HEDL P-19 irradiation experiment.
This paper aims at giving a general overview of the current state of partitioning and transmutation (P&T) studies for advanced nuclear fuel cycles across the world. P&T usually refers to the entire fuel cycle steps linked with minor actinides transmutation, from separation of the minor actinides from the spent fuel to manufacturing of new fuels suited for transmutation, irradiation, handling and reprocessing or disposal of the spent transmutation fuels. The feasibility of most of these steps has been demonstrated in the laboratory scale since the beginning of the nuclear era, however the physical-chemical and radiological properties of minor actinides make their handling considerably more complicated than standard UOX or MOX fuel. As such, research on this topic is still ongoing around the world.
Due to the shutdown of various experimental installations since the 90′ (EBR-II in the US in 1994, PFR in the UK in the same year, Joyo in Japan in 2007, Phénix and Osiris in France in 2009 and 2016 respectively), few irradiation experiments have been carried out in the past decade. The MARIOS and DIAMINO experiments were carried out in the HFR and OSIRIS reactor and showed that the irradiation of fuel with a micro-structure optimized for helium release (and thus limited swelling) was achievable. The irradiation of metallic and oxide containing minor actinides and/or rare earths was carried out in the ATR reactor in the US. The good behaviour of oxide fuel containing minor actinides was demonstrated, although pin failures occurred with the metallic fuel. Furthermore, Post-Irradiation Examinations of various irradiation experiments have been carried out, the most salient one being the METAPHIX experiment in Phenix, which demonstrated the safe behaviour under irradiation of the base alloy considered in the experiment.
Nevertheless, encouraging results in the field of americium separation from the spent fuel have been obtained, the main one being the successful demonstration of the ExAm process to isolate americium from a PUREX raffinate. Furthermore, progress has been made on the manufacturing of UAmO2 targets during the preparation of DIAMINO and MARIOS experiments. Finally, interest in minor actinides transmutation has started to rise again, notably linked with the current trend in molten salt reactor, which are suitable for minor actinides transmutation.
