Fast-neutron spectrum nuclear reactors allow generating carbon-free energy from fissile uranium and plutonium isotopes with increased fuel utilisation compared to currently used light-water reactors (LWRs), whereby they contribute to closing the nuclear fuel cycle. The increased fuel utilisation influences the chemistry of the fuel pin, leading among others to the formation of the so-called JOG-layer (joint oxyde-gaine, French for the connecting layer between the oxide fuel and cladding material). This JOG-phase is chemically approximated as Cs2MoO4. Unlike conventional LWRs, liquid metals are used as coolant to accommodate the fast-neutron spectrum. In this work, liquid lead (Pb) and lead-bismuth eutectic (LBE), a liquid mixture of lead and bismuth (Bi) are considered. The class of reactors using these coolants and a fast neutron spectrum is known as Lead-cooled Fast Reactors (LFRs). This dissertation studies chemical interactions that can occur between coolant and fission products following cladding failure in LFRs, first focussing on the interaction between coolant and JOG-layer. Other important fission products to assess are cesium (Cs) and iodine (I), so-called volatile fission products and barium (Ba), present in the so-called grey phase. It uses the chemical thermodynamics approach, as complementary to post-irradiation examinations and kinetic and release studies.

Chemical thermodynamics centres around a proper description of the Gibbs energy of the possible phases. This Gibbs energy description is informed by available experimental data. Important compounds are typically synthesised using solid state synthesis. Their characterisation involves X-ray and neutron diffraction at ambient and nonambient temperatures, along with X-ray absorption spectroscopy. After the characterisation, thermodynamic properties like the enthalpy of formation and standard entropy are determined. Phase diagrams are measured using differential scanning calorimetry to study phase transition points and know the aggregation state of mixtures in (composition, temperature)-space. The acquired thermodynamic data are used to perform thermochemical calculations or to develop thermodynamic models using the so-called CALPHAD approach.

In this work, the possibility of chemical interaction between Pb-coolant and JOG-layer was studied. Thermodynamic properties of the compounds PbMoO4, Pb2MoO5 and Cs2Pb(MoO4)2, such as standard entropy, enthalpy of formation and melting enthalpy were determined experimentally. Based on this, a complete thermodynamic model of the Pb-Mo-O system, including PbMoO4, Pb2MoO5 and Pb5MoO8 was developed using computational thermochemical software (ThermoCalc). Finally, thermodynamic calculations show that Cs2Pb(MoO4)2, PbMoO4, Pb2MoO5 and Pb5MoO8 can form in LFR operating conditions i.e. with typical oxygen concentrations present in the coolant. Next to this, thermal expansion and Mo-oxidation state ofPbMoO4, Pb2MoO5 and Cs2Pb(MoO4)2 were measured, in order to for example assess the mechanical interaction of these phases after formation.

The compound CsBi(MoO4)2 was studied as a possible formation product between LBE and JOG-phase. A long-standing issue in the understanding of the crystal structure of this compound has been solved using neutron diffraction. The thermal expansion of CsBi(MoO4)2 was determined.

To assess the interaction between coolant and volatile fission products, the system CsI-PbI2-BiI3 was studied experimentally. The low-temperature heat capacity of the three compounds in the system(CsPbI3, Cs4PbI6 and Cs3Bi2I9) were determined and the standard entropy was calculated. The phase diagrams CsI-PbI2, CsI-BiI3 and PbI2-BiI3 were measured using differential scanning calorimetry. A thermodynamic model was developed to predict the liquidus surface of the CsI-PbI2-BiI3 system. The accuracy of the model was confirmed by selective measurements of the ternary eutectics and the pseudo-binary CsPbI3-Cs3Bi2I9.

Study of the interaction between the grey-phase element Ba, fuel and coolant was initiated. During this work, a BaO-deficient plutonium-based perovskite with a composition close to Ba3PuO6 was synthesised. Its crystal structure was studied, as well as the phase transitions at high temperature. The standard entropy of the compound and magnetic susceptibility were determined experimentally. This work, valuable in itself as a contribution to the understanding of the irradiated nuclear fuel pin, is needed as a building block to study coolant-grey phase interaction.

Overall, this thesis describes potential chemical interaction products in the scenario of cladding failure in LFRs. In the concluding chapter, it is shown that the oxygen concentration present in operating conditions allows for the formation of several complex oxide compounds in case of Pb-JOG interaction. In general, this work provides new and necessary data to assess the stability of iodide and oxide compounds. The results present in this thesis should be combined with post-irradiation examination and kinetic studies to assess the scenario of cladding failure from different perspectives.

This thesis investigates the thermochemical behaviour and safety of chloride-based molten salt reactor (MSR) fuels, with a particular focus on the role of fission products during reactor operation. Although molten salt technology was first explored in the 1950s at Oak Ridge National Laboratory through the Aircraft Reactor Experiment and Molten Salt Reactor Experiment, interest declined in favour of other reactor types. Today, however, renewed attention has emerged due to the potential safety and efficiency advantages of molten salt systems, especially their low vapor pressure, high heat capacity, and favourable transport properties.

Modern MSR designs increasingly consider chloride-based fuels instead of traditional fluoride systems, as chlorides allow higher actinide solubility and compatibility with existing reprocessing technologies. However, a major challenge in assessing their safety lies in understanding the behaviour of fission products within these complex, multi-component molten salt systems. These fission products can exist as dissolved species, volatile compounds, or solid precipitates, each posing different operational risks such as precipitation or vaporisation.

The research focuses on both salt-soluble fission products (e.g. barium, strontium, rare earth elements) and volatile species such as cesium and iodine. Their interactions with molten salt fuels are studied experimentally using techniques including Differential Scanning Calorimetry, X-ray diffraction, and neutron diffraction. These experiments are complemented by thermodynamic modelling based on the CALPHAD method, which enables prediction of phase stability and system behaviour under various temperatures and compositions. Central to this modelling is the Gibbs free energy, which determines the most stable phase configurations.

Due to the hazardous nature and limited availability of plutonium chloride (PuCl₃), the study employs simulant materials such as neodymium chloride (NdCl₃) and cerium chloride (CeCl₃), which closely mimic the thermochemical behaviour of plutonium and uranium chlorides. These simulants allow safe experimental investigation while maintaining scientific relevance. The work establishes a thermodynamic description of base fuel systems such as NaCl–MgCl₂–PuCl₃ and validates the use of simulants in representing real reactor conditions.

Significant findings include the identification of previously unknown solid solutions and intermediate compounds in systems containing barium and strontium, improving the understanding of precipitation risks. The research also demonstrates that while individual fission products can increase the likelihood of solid formation, realistic mixtures of fission products—reflecting actual reactor conditions—do not significantly alter the melting behaviour due to their lower concentrations.

The study further explores volatile fission products, particularly iodine and cesium, which may contribute to vaporisation risks. Modelling of complex mixed cation–anion systems shows that chloride-based fuels exhibit stronger retention of these volatile species compared to fluoride-based systems, indicating a safety advantage.

Finally, the developed thermodynamic database is validated against experimental data and applied to simulate real reactor conditions. Results confirm its reliability in predicting phase equilibria and assessing risks related to precipitation and vaporisation. Overall, the thesis provides a comprehensive thermodynamic framework for evaluating the safety of molten salt fuels under irradiation, while also identifying remaining knowledge gaps, such as the need for further experimental validation of certain compounds and interactions.

In conclusion, this work significantly advances the understanding of fission product behaviour in chloride-based molten salt reactors and supports their development as a safe and promising nuclear energy technology.

Safety assessments are the main pillars of the analysis of the impact of storage of the spent nuclear fuel. There are many scenarios to describe what might happen during the storage and disposal time of the nuclear waste. Even though the main composition of the spent nuclear is UO2, the matrix contains transuranium elements and fission products, which have different chemical behaviour and lead to an altered physical state after the irradiation. Thus, the complex nature of the spent nuclear fuels requires understanding of several mechanisms through investigation of individual parameters and their effect on one and another. This is achieved by single effect studies, starting from simple systems to gradually more complex systems. In this thesis, thin films have been used as model systems to simulate the spent fuel in a systematic manner. The main focus was given to the actinide (mixed) oxides (Th, U, Np, Pu, and Ce as surrogate for Pu and as fission product). Throughout this thesis, the suitability of the use of thin films instead of bulk material has been demonstrated, and the investigation of redox properties of model systems for spent fuels using different methods is described.