A new system for the selective separation of Am from high-level liquid waste was developed, composed of only C, H, O, and N atoms. This is a significant step forward from the previous non-CHON system as it only has gaseous combustion products, which reduces secondary waste. The new system handles nominal americium concentrations. Furthermore, distribution data for some key fission products were determined. Finally, the previous non-CHON system’s behaviour under irradiation was studied in detail. We worked on the AmSEL process to recover Am from PUREX raffinate using the new system. The process was successfully tested on a laboratory-scale spiked centrifugal contactor test. The results were analysed and used to adapt the model created earlier to simulate the concentration profiles of Am and curium at the end of the test. We then used the new model design a new optimised flowsheet for a second test with macro-concentrations of Am and curium (Cm) at NNL (UK). The solvent loading was done using the methodology from the first test to produce a very similar flowsheet feed as before to aid comparison. Analysis of the results is in the final stages.
Next, we studied how the composition of AmSel stripping solutions affects the way to make Am bearing fuel by a powderless process. For that we made and characterised 3-30% Nd doped ZrO2 inert matrix microspheres by internal gelation and powder by coprecipitation. In parallel we looked into the impact of the AmSel system on UAmO2 co-conversion with oxalic precipitation and investigated Neptunium (Np) via the dissolution behaviour of NpO2 and hydrolysis of Np ions under aqueous conditions relevant for internal gelation methods. We also produced porous uranium oxides microspheres by internal gelation process, calcination and finally infiltration with a Nd solution to form after calcination and sintering, U1-yNdyO2-x beads.
A second main goal is to improve fuel performance codes (FPC) as applied to Am bearing fuel. We did detailed examinations of irradiated (U,Am)O2 fuel pins and experimental measurements as well as atomistic simulations of mixed oxides stable in the U-Am-Np-Pu-O system and used the data to develop and improve thermodynamic models of the oxides of Pu and U combined with Am and Np based on the CALPHAD methodology. A database of the results was established. We then validated the improved FPC by comparing simulations of earlier irradiation tests with experiment. The FPC’s were also applied to investigate the behaviour of Am-bearing fuels in a transmutation pin in MYRRHA. For normal working conditions we considered mixed oxide fuel with an Am content from 0.49 to 5% and evaluated the safety criteria and their sensitivity to the Am content. For off-normal conditions we looked to the most severe transient, i.e. a beam power jump when the proton beam switches on abruptly. We also did preliminary designs of future transient tests in the HFR and BR2, aiming to assess the most relevant scenarios for reactor safety, and evaluated other facilities for transient testing with Am-bearing fuels to identify those that can achieve faster transient kinetics, as those achievable in material test reactors.
In the rest of the project we focus on the reactor core and system safety of MYYRHA, testing material behaviour and chemical interactions in the lead-bismuth eutectic (LBE) coolant by exposing a heated fuel bundle mock-up for six months in representative conditions validating thermal-hydraulic and chemistry models. This showed no plugging risks from corrosion products and a pressure drop only occuring above PbO precipitation. Moreover, the mechanical properties of corroded cladding material showed no degradation beyond the corroded layer.
A detailed analyses of pin segments subjected to a fast transient irradiation in the MAXIMA project, showed no cracks, even in the zones with the largest plastic deformation caused by the pellet-cladding interaction. A thermal analysis of the test using both the ANSYS and TRANSURANUS codes yielded very similar results which were used in a structural analysis using ANSYS. The results agreed very well with experimental data.
The final action for core safety is an experimental and numercal study of blocked and bent rod bundles. For a partially blocked rod bundle with a well defined porosity, measurement and simulation showed a moderate local overheating of the rods. The bent-rod experiments was set up and preliminary calculations were carried out.
Finally, working on the safety of the MYRRHA Accelerator-Driven System, data from comparable accelerators showed that to reach the needed reliability, tests and fault tolerance schemes on the MINERVA accelerator should be envisaged. In studies on beam window breakage and radionuclide release we were able to detect evaporated mercury by a mass spectrometer added at the end of the stainless-steel pipe representative of the beam line. Further analysis is needed to understand the observed mechanisms of evaporation and transport. The POVACS setup, to investigate polonium sticking on stainless steel has been constructed, tested with LBE, and nearly licensed. Experiments have been planned.
On pool thermal hydrualics, the work done was relatated mainly to the experiment and simulations on ESCAPE and CIRCE THETIS facilities. The experiments in ESCAPE was carried out while the experiment in the CIRCE facility will done out in the next months. The numerical activities related to the experiments were mainly carried out regarding the pre-test, while for the post-test in CIRCE will be done in 2025. The AHFM was developed and validated.
We also did experiments and calculations to study the chemical processes in the LBE coolant looking at radioactivity release, essential for MYRRHA's safety assessment, progressed in understanding the release behaviour of polonium, got experimental results for the release of tellurium from LBE and its speciation. Finally, the stability of associated molecules was calculated using computational chemistry.
