[en] Molten Salt Reactors (MSRs) are rather category of reactors than a single concept. In majority of the concepts the fuel cycle performance profits from the consideration that selected gaseous, volatile, non-soluble fission products could be removed during the operation from the liquid fuel. Many concepts also foresee integration of fuel cleaning, or actually processing, unit in the same complex. Accordingly, the distribution or radiotoxicity in the MSR system may strongly differ from classical reactors with solid fuel. At the same time, the presence of driving forces can be eliminated.

[en] Highlights: • Coupling of a sever accident code with a thermodynamic modelling package. • Improved description of the Mo and Ba species release from the nuclear fuel. • The developed code reproduced fission products release during the VERDON experiment. • The species partitioning and release behaviour depends on the redox conditions. The treatment of chemistry and thermodynamics in the integral severe accident codes is typically limited. A more accurate treatment of the chemistry during the severe accident modelling is, therefore, of great interest. For this purpose, the work is focused on the development of coupling of the MELCOR code with chemical thermodynamic calculations using GEMS codes and HERACLES database. Developed coupling between the two codes, called cGEMS, allowed for the improved thermodynamic description of the fission product release from the nuclear fuel under severe accident conditions. VERDON-1 test was selected as an experimental reference for the simulations. Experimental release behaviour of Mo, Cs and Ba observed in VERDON-1 experiment was reproduced by the developed coupled code. Performed simulations provided detailed information about the fission product speciation at different redox conditions. The obtained information and the developed code provides a more accurate description of the fission product behaviour and release during severe accidents.

[en] The different approaches available for the estimation of the reactor irradiation effect on the thermal properties of oxide fuels are reviewed, from the experimental determinations to atomistic techniques. (author)

[en] In the present work, calculations of the thermal conductivity of UO₂ were carried out applying classical Molecular Dynamics for the isothermal-isobaric (NPT) statistical ensemble, using the Green–Kubo approach. The thermal conductivity calculated for perfect stoichiometric UO₂ is in good agreement with the literature data over the temperature range corresponding to heat transfer by phonons (up to 1700 K). The effect of non-stoichiometry on the thermal conductivity was calculated taking into account the presence of polarons. It was found that for the same value of the stoichiometry deviation, the effect of oxygen vacancies (hypo-stoichiometry) is more pronounced than the effect of oxygen interstitials (hyper-stoichiometry). Then the influence of the oxygen Frenkel pairs on the thermal conductivity was calculated. The simultaneous impact of non-stoichiometry and OFP on the thermal conductivity was investigated and it was shown that the two effects can be combined using the interpretation obtained with the classical phonons scattering theory. Finally, simplified correlations were deduced for the calculation of the thermal conductivity of UO₂ taking into account the effect of non-stoichiometry and of Frenkel pairs, these two effects being present during irradiation

[en] This article presents a synthesis of the investigations at the atomic scale of the transport properties of defects and fission gases in uranium dioxide, as well as of the transfer of results from the atomic scale to models at the mesoscopic scale, performed during the F-BRIDGE European project (2008–2012). We first present the mesoscale models used to investigate uranium oxide fuel under irradiation, and in particular the cluster dynamics and kinetic Monte Carlo methods employed to model the behaviour of defects and fission gases in UO_2, as well as the parameters of these models. Second, we describe briefly the atomic scale methods employed, i.e. electronic structure calculations and empirical potential methods. Then, we show the results of the calculation of the data necessary for the mesoscale models using these atomic scale methods. Finally, we summarise the links built between the atomic and mesoscopic scale by listing the data calculated at the atomic scale which are to be used as input in mesoscale modelling. Despite specific difficulties in the description of fuel materials, the results obtained in F-BRIDGE show that atomic scale modelling methods are now mature enough to obtain precise data to feed higher scale models and help interpret experiments on nuclear fuels. These methods bring valuable insight, in particular the formation, binding and migration energies of point and extended defects, fission product localization, incorporation energies and migration pathways, elementary mechanisms of irradiation induced processes. These studies open the way for the investigation of other significant phenomena involved in fuel behaviour, in particular the thermochemical and thermomechanical properties and their evolution in-pile, complex microstructures, as well as of more complex fuels