The aim of this paper is to present methods for corrosion mitigation in molten salt environments. The corrosion of structural materials depends directly on the redox potential of the salt. When the redox potential of the salt is higher than the standard potentials of the elements constituting the structural materials, corrosion occurs. If the reverse is true, no corrosion is observed. Herein, a methodology for calculating the theoretical potential of a molten salt is provided and compared with experimental measurements. Three ways to mitigate corrosion by modifying the salt redox potential are proposed: (i) using a soluble/soluble redox system; (ii) using a potentiostatic method; and (iii) using an amphoteric compound such as UCl₃, TiCl₂, or TiCl₃. Immersion tests were conducted under the above conditions to validate the methodology.

Reliable neutron-induced-reaction cross sections of unstable nuclei are essential for nuclear astrophysics and applications but their direct measurement is often impossible. The surrogate-reaction method is one of the most promising alternatives to access these cross sections. In this work, we successfully applied the surrogate-reaction method to infer for the first time both the neutron-induced fission and radiative capture cross sections of ${}^{239}\mathrm{Pu}$ in a consistent manner from a single measurement. This was achieved by combining simultaneously measured fission and $\gamma$-emission probabilities for the ${}^{240}\mathrm{Pu}({}^4\mathrm{He},{}^4{\mathrm{He}}^{\prime })$ surrogate reaction with a calculation of the angular-momentum and parity distributions populated in this reaction. While other experiments measure the probabilities for some selected $\gamma$-ray transitions, we measure the $\gamma$-emission probability. This enlarges the applicability of the surrogate-reaction method.

The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR) concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF₄-UF₄-NdF₃ with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

Aluminum reactivity in cement-based matrices, generally used for conditioning low- and intermediate-level radioactive wastes, is a problem due to dihydrogen produced by the corrosion reaction which depends on the pH of the pore solution. Electrochemical impedance spectroscopy has been used to propose a mechanism for aluminum corrosion in cementitious matrices based on ordinary Portland cement (OPC, pH ≈ 13) or magnesium phosphate cement (MKP, pH ranging between 4 and 9) containing or not LiNO₃ as a corrosion inhibitor. The fit of impedance diagrams recorded on aluminum electrode as a function of time has been realized using electrical parameters to model the cement and kinetic constants for the faradaic impedance. The so determined kinetic constants are used to calculate the corrosion current and the dihydrogen production as a function of time. Comparison of these results with experimental measurements of dihydrogen release obtained by gas chromatography shows a very good agreement except in the MKP matrix containing LiNO₃. In the OPC matrix, dihydrogen production is between 500 and 1000 times higher than in the MKP matrix, which put in evidence the significant benefit in using MKP cements for aluminum encapsulation.

Portland cement is extensively used for the conditioning of radioactive waste. However, its high alkalinity is a serious obstacle to the stabilization of waste containing aluminum metal since aluminum is oxidized by the pore solution with the production of dihydrogen. This work investigates the potential of an alternative binder, magnesium potassium phosphate (MKP) cement, for the stabilization of Al–Mg alloys comprising 2 to 4.5 wt% of Mg and other metallic impurities. The objective is to assess the influence of the alloy composition on its reactivity in the cementitious matrix at earlier ages, as well as at later ages, when the cement has reached a significant reaction degree. Two complementary techniques are used. Gas chromatography shows that the dihydrogen release, resulting from the corrosion process, is not influenced by the magnesium content in the alloy. Electrochemical impedance spectroscopy provides qualitative information about the corrosion but also makes it possible to assess the corrosion current using an equivalent electrical circuit linked to the kinetic parameters of the postulated corrosion mechanism. Over a one-year period, the corrosion current of the alloys, regardless of their Mg content, is reduced by almost three orders of magnitude in MKP mortar as compared to Portland-cement-based mortar.

Thorium oxide is poorly soluble: unlike uranium oxide, concentrated nitric acid medium is not sufficient to get quantitative dissolution. Addition of small amounts of fluoride is required to achieve thorium oxide total dissolution. The effect of several parameters on thorium oxide dissolution in order to optimize the dissolution conditions is reported in this paper. Thus the influence of solid characteristics, dissolution method, temperature and composition of dissolution medium on ThO₂dissolution rate has been studied. No complexing agents tested other than fluoride allows total dissolution. Beyond a given HF concentration a decrease of the dissolution rate is observed due to the formation of a precipitate at the solid/solution interface. It was demonstrated by XPS measurements that this precipitate is constituted of thorium fluoride (ThF₄) formed during the ThO₂dissolution. The low concentration of HF required to achieve a total dissolution and the activation energy value measured tends to show a catalytic effect of HF on the dissolution process.