O. Beneš, P.R. Hania, E. D’Agata, A. Rodrigues, H.J. Uitslag-Doolaard, R. Konings.
European Commission, Joint Research Centre (JRC), Karlsruhe, Germany, Nuclear Research; Consultancy Group, P.O. Box 25 1755 ZG Petten, The Netherlands, European Commission, Joint Research Centre (JRC), Petten, The Netherlands.
In many of the studied Molten Salt Reactor (MSR) concepts, molten salt is used as a carrier media for the fuel compounds dissolved in the salt, whereas the salt serves at the same time as a primary coolant of the reactor and as neutron moderator. The interest for MSR technologies is increasing worldwide not only within national programmes, e.g. in France, USA, China and others, but different concepts are being developed commercially by several start-up companies, e.g., Terrestrial Energy (Canada/US), TerraPower (USA), Moltex (UK), Seaborg Technologies (Denmark) and others. In addition, MSR is one of the advanced reactor concepts studied within the Generation IV Forum since 2002. The main advantage of this type of reactor is its enhanced safety with its strong negative temperature coefficient of reactivity and improved sustainability. The MSR is operated at low (atmospheric) pressure in the primary circuit and in the event of overheating, the fuel would be drained into an emergency dump tank assuring subcriticality and natural removal of the decay heat.
The irradiation experiment SALIENT-03 is being carried out within collaboration between the Nuclear Research and Consultancy Group (NRG) and the Joint Research Centre (JRC). The main goal of the experiment is to assess the corrosion mechanism of selected Ni-based alloys in molten fluoride salt considered as one of the candidates for MSR fuel salt. The corrosion test will be carried out during irradiation of the fuel salt in the High Flux Reactor in Petten (The Netherlands). A molten salt based on the 78LiF-22ThF4 eutectic mixture was selected, as it is a carrier salt for the Molten Salt Fast Reactor (MSFR) concept studied in Europe. The MSFR is based on a non-moderated neutron spectrum utilising the 232Th-233U fuel cycle: it is being developed in France and since 2015, its safety assessment is being evaluated within the coordinated R&D projects in the frame of the EC/EURATOM programmes. In the SALIENT-03 experiment, the fissile material will be dissolved in this carrier melt to form a fuel salt with the composition of 75.0LiF-18.7ThF4-6.0UF4/UF3-0.3PuF3.
The present work is describing the synthesis and characterisation of pure actinide fluorides needed for the experiment, i.e., ThF4, UF4, UF3 and PuF3. Since the details on the syntheses of ThF4, UF4 and PuF3 have been already published, the main focus is dedicated to the preparation of UF3. This compound is needed not only to form the desired UF4/UF3 ratio in the irradiated molten salt, but also to calibrate an electrochemical method for the detection of this ratio in the fuel. Since it is difficult to analytically determine the ratio in the mixture, the work aimed to synthesis of phase pure UF3. The method had to be thoroughly optimised to prevent disproportionation of UF3 to U metal and UF4 and/or formation of oxides and to achieve a product purity of 99.6%.
Event Timeslots (1)
Wednesday – 15th September 2021