This study establishes an in-situ synchrotron approach to directly link molten-salt chemistry, redox-buffer speciation, and alloy corrosion in chloride melts. Using a first-of-its-kind corrosion cell coupled to simultaneous X-ray absorption spectroscopy (XAS) and X-ray fluorescence (XRF), we monitor europium redox-buffer speciation and chromium dissolution during corrosion of Ni-20Cr in molten LiCl-KCl and KCl-MgCl 2 salts. To build a mechanistic framework, we combine UV-Vis optical basicity measurements, in-situ electrochemistry, and molecular dynamics simulations to quantify how salt chlorobasicity reshapes europium solvation and shifts the Eu(III)/Eu(II) apparent potential. Increasing MgCl 2 content drives the apparent potential of the Eu(III)/Eu(II) couple +0.45 V relative to LiCl-KCl and stabilizes Eu(II)-compatible solvation environments. In-situ, time-resolved Eu L3-edge XANES reveals progressive Eu(III) reduction in KCl-MgCl 2 melts, while Eu(III) remains predominantly stable in LiCl-KCl. This behavior is consistent with a chlorobasicity-activated reaction with boron nitride (BN) cell components. In less chlorobasic melts, salt redox chemistry enables BN chlorination, allowing BN to act as a sacrificial reductant and chlorinating agent that promotes Eu(III) to Eu(II) conversion, whereas this pathway is not thermodynamically favored in LiCl-KCl. During active corrosion of Ni-20Cr, coupled Eu XANES and Cr XRF measurements show concurrent Eu(III) to Eu(II) conversion and Cr dissolution, confirming that Eu(III) is the oxidizing species driving chromium oxidation. Together, these results demonstrate that chlorobasicity is a central control variable for redox-buffer speciation and corrosion thermodynamics, and more broadly that chlorobasicity can enable or suppress reactive pathways with structural materials in high-temperature molten salts.

This study focuses on mitigating corrosion in the molten chloride salt NaCl-MgCl2 by stabilizing the salt redox potential at similar or equal to-2 vs (Cl2/Cl-)/V through the use of the SmCl3/SmCl2 redox couple, combined with oxoacidity regulation via AlCl3. The reversible, thermally stable and soluble/soluble SmCl3/SmCl2 system effectively imposes a stable potential of-2 V at 700 degrees C. Aluminum plays a dual role: it reduces SmCl3 to SmCl2 and regulates oxoacidity, preventing samarium precipitation and ensuring stability of the x(SmCl3)/x(SmCl2) (ratio 1:5), and consequently the redox potential. Corrosion tests were conducted on Ni-based (Alloy 625) and Fe-based (316 L stainless steel) alloys immersed in NaCl-MgCl2 at 700 degrees C for 13 days. The salt chemistry was controlled by fixing both the oxoacidity and the redox potential. The results demonstrate the effectiveness of this approach by positioning metallic materials within their immunity domain. Alloy 625 exhibited no intergranular neither micrometric-scale corrosion, whereas 316 L stainless steel showed only a 1 & micro;m corrosion layer. This is a significant improvement compared to the loss of 100 & micro;m of material observed when the salt chemistry was not controlled. The redox potential remained stable at-1.96 vs (Cl2/Cl-)/V, confirming the robustness of the redox potential and oxoacidity control approach.

Molten Salt Reactors (MSR) are Generation IV nuclear systems in which the fuel is dissolved in a molten salt circulating through the primary system. There is growing interest in this advanced technology in Europe, but also in the US, China, South Korea, Japan and Russia, due to their inherently high safety level, flexibility, reliability, load-following capabilities, and potential for multi-recycling of materials contained in light-water reactors’ spent nuclear fuels. These advantages could position MSRs as ideal complements to other decarbonized energy sources in a future sustainable energy mix. In this respect, it is probably one of the most promising advanced technologies and, at the same time, the least mature and studied one. Two ongoing EURATOM-funded projects, MIMOSA and ENDURANCE projects are exploring molten salt reactors’ safety and performance features, as well as fuel cycle aspects, in order to assess and demonstrate their potential for future deployment in Europe. The MIMOSA and ENDURANCE projects have the common objective of improving the maturity of MSR technology. The MIMOSA project develops and analyses multi-recycling strategies for the European Union based on the use of MSR and demonstrates several key aspects of their technical feasibility and performance by both calculations and experimental investigations. The ENDURANCE project supports the safe operation and the development of Critical Technology Elements by connecting design developers and industry with universities and research centres while ensuring alignment with regulatory requirements. Whereas ENDURANCE is in its starting phase, MIMOSA has already delivered important results.

New nuclear technologies could involve the extensive use of molten salts, including actinide halides. Despite their importance, several practical challenges limit experimental measurements, resulting in knowledge gaps for structural and thermophysical properties. In this work, new polarizable force fields based on ab initio calculations for the simulation of molten actinide chlorides are introduced. The new force fields are used to compute structural properties, density, heat capacity, and isothermal compressibility of pure actinide molten salts (ThCl4, PaCl3, NpCl3, AmCl3, CmCl3) at various temperatures. UCl3 and PuCl3, which were parameterized in previous works, are also included. The results are discussed in the context of already existing theoretical and experimental datasets, showing good agreement with the literature. Predictions are extended to systems not considered in previous works. Notably, the results highlight the peculiarity of ThCl4 compared to actinide trichlorides in terms of structural and thermophysical properties. The new force fields can be used in future works for the simulation of molten salts mixtures containing actinides.

This study examines a thermodynamic approach to mitigating metal corrosion in molten chloride salts, focusing on lowering the salt's potential to protect structural materials. Thermodynamic diagrams show that the constituents of stainless steel and Ni-based alloys, i.e. Fe, Ni, and Cr, are susceptible to oxidation due to their open circuit potential (OCP) relative to the potential of the salt. Mo is an exception, as it is inert. The addition of TiCl2 to the molten salt effectively reduces its potential, buffering it at- 1.07 V, below the oxidation threshold of Cr. This reduction is achieved by in situ synthesis of TiCl2 by ZnCl2, which prevents excessive oxidation of Ti to higher states, and maintains the stability of the salt potential over time. Corrosion tests on 304 L stainless steel and Inconel (R) 600 show significant corrosion without TiCl2, as the unbuffered salt potential exceed the immunity domain of these alloys. However, the addition of TiCl2 shifts the salt potential into the immunity domain of Fe, Ni, and Cr, preventing corrosion. The results that a controlled, lower salt potential is crucial for ensuring material stability in molten chloride environments. The results highlight the importance of maintaining a zero redox potential difference between the salt and the metals to achieve inertness. Therefore, controlling the molten salt potential appears to be essential when considering structural integrity in harsh molten salt environments.

Metallic beryllium is an attractive material for nuclear applications, e.g. thermal reactor and fusion reactors. After use, activated beryllium waste will need to be disposed of in a geological disposal repository. One option for the treatment and conditioning of the beryllium waste could be the direct emplacement of pieces of beryllium in a cementitious matrix. In these conditions, beryllium corrodes due to reaction with water to produce hydrogen, and beryllium hydroxide, leading to possible loss of the waste containment. To gain more information on the anaerobic corrosion of beryllium, Electrochemical Impedance Spectroscopy and hydrogen measurement by gas chromatography were used, while beryllium samples were immersed in solutions representative of promising matrices (Portland cement (OPC, pH 13.5) and magnesium phosphate cement (MPC, pH 8.1)) for maximum one year. At the beginning of the test (< 180 days), both techniques showed that the beryllium corrosion rate was lower in OPC solution than in MPC solution. At longer term (> 180 days), the corrosion rate decreased sharply in MPC solution, while the corrosion rate in OPC solution remained stable. Therefore, at longer term, the lower corrosion rate was observed in MPC solution. The reason of this corrosion rate decrease comes from the formation of a passive layer composed of Be(OH)(2) (inner part) and KBePO4H2O crystals (outer part) at the surface of beryllium in MPC solution, while no protective corrosion product layer was present on the metal surface in OPC solution. In that later case, pitting corrosion was observed in addition to the dissolution of beryllium species in solution probably in the form of Be(OH)(4)(2-).

Thanks to species identification and growth interface localization experiments, a mechanism of iron corrosion in liquid UF6 at 80 degrees C was suggested. After an UF6 dissociation step, resulting fluorine (HF, F- or F center dot) was adsorbed at the FeF2 external interface and diffused into the iron fluoride layer via fluorine vacancy. FeF2 grew thus at the FeF2/Fe interface. Two FeF2 growth kinetics were observed depending on the presence or absence of NOxF impurities. Both corrosion rates were controlled by the cathodic reaction and presence of NOxF catalysed it.

Magnesium potassium phosphate cement (MKPC) is a promising material for encapsulating low-to intermediate-level radioactive metallic wastes, including aluminium. This study investigates the influence of magnesium oxide (MgO) type and chemical retarders-boric acid and sodium thiosulfate-on the passivation of aluminium A1050 alloy. Corrosion behaviour and hydrogen evolution were monitored over 25 days in aqueous solutions and 230 days in mortars using electrochemical impedance spectroscopy and linear polarisation resistance. In aqueous solutions, boric acid improved passivation by shifting the corrosion potential to more anodic values, while thiosulfate promoted rapid passivation through adsorption, leading to slightly lower anodic potentials. In mortars, the near-neutral pH facilitated passivation, though variations were observed depending on the MgO type and retarder composition. Soft-burnt MgO combined with thiosulfate accelerated the formation of the alumina layer, enhancing passivation and reducing corrosion kinetics due to its higher reactivity and refined pore structure. In contrast, hard-burnt MgO with boric acid resulted in slower passivation and more cathodic potentials, likely due to its lower reactivity and heterogeneous pore network. When both retarders were used together, a competitive effect was observed, reducing passivation efficiency and increasing hydrogen evolution despite anodic potential shifts. These findings underscore the critical role of MgO type and retarder selection in optimising aluminium passivation within MKPC systems.

The results reported here are the first on pure iron corrosion in liquid UF6 at 80 degrees C. Two kinetic behaviours have been observed: one led to micrometric scales (from micron to hundreds of microns after several months) and the other one to hundreds of nanometers for several months. The higher corrosion kinetics resulted in the presence of impurities such as NOxF complexes formed by interaction between the medium and the reactor material. These NOxF catalysed the corrosion reaction leading to a corrosion mechanism controlled by the cathodic reaction rate. Effect of impurities in UF6 coming from experimental conditions or nature of uranium ore should then be systematically and carefully checked. Whatever the corrosion kinetics and the presence of impurities, the nature of the layer was identical: a duplex fluoride scale composed of an iron rich layer, FeF2, and a uranium rich layer evolving over time from U2F9 to UF5.

Beryllium metal is used as neutron moderator and reflector or multiplier in certain types of fission or fusion reactors. Dismantling of these reactors will produce radioactive beryllium waste, classified as low- or intermediate-level waste, that will need to be stabilised and solidified before being sent to disposal. The cementation process is under consideration because it may offer a good compromise between simplicity of implementation, cost, and quality of the final cemented wasteform. Nevertheless, knowledge of the corrosion behaviour of Be metal in a cement-based matrix is still limited, partly due to the high toxicity of Be that complicates testing. This study thus investigates Be corrosion in cement suspensions using potentiometry, voltammetry, and electrochemical impedance spectroscopy. Among the five different investigated systems (Portland cement blended without or with 40 wt.% silica fume, calcium sulfoaluminate clinker blended without or with 15% anhydrite, and calcium aluminate cement), Portland cement blended with 40% silica fume and calcium sulfoaluminate cement comprising 15% anhydrite are the most effective in mitigating beryllium corrosion. They allow reduction in the corrosion current by factors of 4 and 50, respectively, as compared to Portland cement.

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.

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.

The electrochemical behavior of iodide ions has been studied in the ternary fluoride salt, LiF-NaF-KF and in the binary salts, LiF-CaF2 and LiF-ThF4. The electrochemical study demonstrated that iodide ions are oxidized to produce gaseous species in the three molten fluoride salts. However, the stability of iodide ions was observed to be influenced by the fluroacidity of the molten salt. The efficiency of the extraction of iodide ions was examined in LiF-NaF-KF and LiF-ThF4. UV-visible spectroscopy was used to quantify the amount of iodide ions oxidized after the execution of several coulometries at applied potential, these coulometries simulated the fluorination step in the reprocessing unit designed for the MSFR. The efficiency of extraction determined in LiF-ThF4 is higher than 95 % while it is close to 64 % in LiF-NaF-KF at 650°C.

Beryllium wastes produced by nuclear industry could be managed by encapsulation in cements. The main risk is the aqueous corrosion, which leads to the hydrogen production and cracks causing a loss of radioactivity confinement. However, the corrosion can be limited by the formation of the hydroxide solid phase Be(OH)(2(s)). A study on the influence of the pH on the reactivity of beryllium was carried out in aqueous solution. The solubility diagram was drawn and showed that the hydroxide solid phase is stable until very high basic pH of 14. The pH variation of solution at different initial pH was measured after immersion of metallic beryllium. The results are in good agreement with the calculated solubility diagram. The open circuit potential (OPC) and the electrochemical impedance spectra (EIS) were recorded at a beryllium electrode from pH 2 to 15. The simulation of the EIS by using an electric equivalent circuit allowed to evaluate the Be reactivity by measuring the charge transfer resistance as a function of the pH. The experimental results confirm that the corrosion of beryllium is limited at neutral to basic pH and this low reactivity was attributed to the formation of the hydroxide solid phase Be(OH)(2). The OCP and EIS measurements put also in evidence that the corrosion of beryllium is the lowest in solution having pH around 12. In this pH range, the Be metal would be protected against the corrosion both by the formation of the hydroxide solid Be(OH)(2) and by the adsorption of hydroxide ions (OH-) at the metal surface.

Liquid-fueled reactors exhibit unusual and interesting properties compared to solid-fueled reactors, requesting a revision of some well-known conception and safety rules. Emphasis is thus put in this chapter on such differences and the need for innovative approaches with the focus on homogeneous molten salt reactors. The molten salt fast reactor concept, where the circulating fuel also plays the role of coolant and based on a fast neutron spectrum, is seen as a long-term alternative to solid-fueled fast reactors. It fulfills the Generation-IV criteria and is studied since a decade mainly by calculations and determination of basic physical and chemical properties in European Union and Russian Federation. The main characteristics of this concept are presented and discussed including transient simulation, chemistry and material issues, safety analysis, and research roadmap and perspectives.

Beryllium wastes are produced by nuclear industry. One way to manage them is their encapsulation in cements. The main risk of this conditioning is the aqueous corrosion, which leads to the hydrogen production and cracks causing a loss of radioactivity confinement. The corrosion can be limited by the formation of the hydroxide solid phase Be(OH)(2(s)). The stability domain of this phase was calculated in water as a function of the pH: M. Pourbaix has calculated a stability domain from 2.9 to 11.7 for a 10(-4) M beryllium concentration, while according to our calculation with more recent thermodynamic data, it is stable from 5.3 to 13.5. Based on Pourbaix results, beryllium cannot be conditioned in the mainly used cement for nuclear waste, Portland cement, while it is possible according to our calculations. Experimental measurements were achieved to select the data set most in agreement with the experimental observations. The beryllium reactivity has been examined in matrices having different pH pore solution: brushite cement (pH 1.75-6.44), magnesium phosphate cement (pH 5.6-8.4), calcium-sulfoaluminate cement (pH 10.9-12.3), Portland cement (pH 12.5-12.9) cements and activated slag (pH 12.9-13.8), by measuring the open circuit potential and by electrochemical impedance spectroscopy. The experimental results agree with the more recent thermodynamic data. Beryllium corrosion is too high in the brushite cement, leading to a high hydrogen production. This matrix can then not be envisaged for the conditioning of Be waste. If the beryllium is encapsulated in the activated slag, the highly alkalinity is too high in the early age, leading to a high aqueous corrosion. Activated slag are also not suitable for Be conditioning. The main conclusion of this paper is that beryllium can be encapsulated in safe conditions in Portland, magnesium phosphate and calcium sulfoaluminate cements. (C) 2021 Elsevier B.V. All rights reserved.

Thermodynamic and electrochemical properties of actinides in molten fluoride salts are one of the key data required for design and safety assessment of any molten salt reactor concept using liquid fluoride based fuel. In the case of the Molten Salt Fast Reactor (MSFR), special attention has to be paid to thorium fluoride, which is a direct fuel carrier salt constituent. This work presents experimental measurement of ThF4 activity coefficient in eutectic LiF-CaF2 melt by electrochemical techniques. The proposed approach is overcoming problems related to the lack of a thermodynamic reference electrode for high-temperature molten fluoride media. The method can be used to compare stability of fissile material and fission products in a given molten fluoride solvent by estimating their activity coefficients using the described electrochemical measurements. In addition, electrochemical methods were successfully used to evaluate and confirm purity of the used materials, with a special regard to oxygen content in the ThF4 input material synthesised during the previous work from ThO2.

The objective of this work is to study the Li2CO3-Na2CO3-K2CO3 (29.5-31.1-39.4 mol%) molten salt and the behaviour of several metals (Au, Pt, C, Ni and W) by electrochemical measurements in the salt under three atmospheres (Ar, CO2 and O-2/N-2). Thermodynamic calculations show that the molten salt anodic and cathodic limits correspond to the carbonate ions oxidation and reduction to dioxygen and carbon respectively. An internal reference electrode has been defined based on the redox system Na+/Au2Na, which is independent of the lithium oxide activity. The widths of the electroactivity domain have been measured on gold working electrode by cyclic voltammetry: 2.21 V under Ar and O-2/N-2 and 2.41 V under CO2, which are larger than the values obtained by thermodynamic calculations (1.280 V and 1.025 V respectively). Gas chromatographic (GC) analysis during electrolysis at potentials ranging in the anodic limit allowed to conclude that the electrochemical domain of the carbonate molten salt is limited anodically by the oxidation of carbonate ions to dioxygen, possibly peroxide ions, and carbon dioxide. The characterisation by XRD of a platinum electrode after electrolysis at the cathodic limit has shown that the electroactivity domain of the molten salt is defined by the carbonate ions reduction into carbon, indicating that CO32-/C is an irreversible system. A carbonate ions decomposition rate of 1 x 10(-3) mmol.h(-1).cm(-2) has been determined by GC and thermogravimetric analysis (TGA) under inert gas. A stabilization of the CO2 pressure is observed after the melting of the salt. (c) 2021 Elsevier Ltd. All rights reserved.