[en] The melted steel is contacted with a prepared slag 1 for chromium oxidation a slag 2 and a metallic alloy A are obtained, they are separated by decantation. The alloy A is oxidized by gaseous oxygen, a slag 3 and an alloy B are obtained. Alloy B has a high Co and/or Ni content and is separated from the slag 3 by decantation

[en] Nuclear fuel elements made of sintered uranium oxide in a metal sheath can trap fission products created during irradiation, if the pellets contain, or are coated, or the sheath is coated inside, by an oxide trapping the fission products

[en] It would often be of benefit to limit the uncertainties involved in the measurements of vapor pressure by the Knudsen effusion method. Major uncertainties, specific to the system investigated or due to parasitic reactions may be suppressed by connecting a mass spectrometer. Nevertheless, some uncertainties persist, due to various causes that can be often identified but remain difficult to assess, and that are specific to the effusion method and the sampling mode. Apart from its analytical function, and due to its speed and dynamic range of measurement, in particular, the mass spectrometer seems particularly well adapted to detect and evaluate any parasitic phenomena that are likely to jeopardize the results

[en] It would often be of benefit to limit the uncertainties involved in the measurements of vapor pressure by the Knudsen effusion method. Major uncertainties, specific to the system investigated or due to parasitic reactions may be suppressed by connecting a mass spectrometer. Nevertheless, some uncertainties persist, due to various causes that can be often identified but remain difficult to assess, and that are specific to the effusion method and the sampling mode. Apart from its analytical function, and due to its speed and dynamic range of measurement, in particular, the mass spectrometer seems particularly well adapted to detect and evaluate any parasitic phenomena that are likely to jeopardize the results

No abstract available

[en] Highlights: • Neutronic calculations for fast spectrum molten salt reactor. • Evaluation of the fissile matter to be used in such reactor as initial fissile load. • Capabilities to transmute transuranic elements. • Deployment scenarios of the Thorium fuel cycle. • Waste management optimization with molten salt fast reactor. - Abstract: There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, ²³³U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs

[en] Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

[en] In the MSFR (Molten Salt Fast Reactor), the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this design characteristic, the MSFR can thus operate with a widely varying fuel composition. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for Bateman's equations computing the population of any nucleus inside any part of the reactor at any moment. The classical Bateman's equations have been modified by adding 2 terms representing the reprocessing capacities and an online addition. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system and radioprotection issues. The very preliminary results show the potential of the neutronic-reprocessing coupling we have developed. We also show that these studies are limited by the uncertainties on the design and the knowledge of the chemical reprocessing processes. (A.C.)

[en] A technology and a protocol were developed for laboratory-scale determination of the distribution coefficients of uranium, representing the actinides and lanthanides. The separation was performed at 720 deg C by selective reduction of actinides fluorides by magnesium. The experimental results show that the distribution of species between the two phases diverges from the ideal situations; this is attributed to infinite-dilution activity coefficients far from 1 for the solutes. Under these conditions, uranium can be separated from lanthanum by carefully controlling the concentration of reducing agent. Regarding kinetics, it seems that the limiting factor is the chemical reaction. (author)

[en] Dismantling power Nuclear plants leads to solve the problem of Cobalt elimination (highly active element) coming from stainless steels. The present study has been carried out in two different ways: ESR (Electro Slag Refining) process application and controlled oxidation with slags. The first one had been given up because of results, at the laboratory scale, confirming thermodynamic calculation according to which Cobalt phosphides (the only compound stabler than Nickel or Iron phosphides) were not enough soluble in slags to be of any interest. The second one consisting of melting stainless steel with slags under a controlled oxidation did not lead to conclusive results, mainly because of too weak chemical reactions rates. Nevertheless, on this basis a project study of an industrial waste treatment plant had been carried-out