Radiolytic fluorine gas production at temperatures of 40°C to 60°C was investigated for the fluoride salts LiF, BeF₂, UF₄, ThF₄, and 71.7LiF-16BeF₂-12.3UF₄ (FliBe-UF₄) by gamma irradiation of powdered samples using spent fuel elements from the High Flux Reactor (HFR) Petten as the irradiation source; work of a similar nature was previously performed at Oak Ridge National Laboratory in the period 1965 to 1995. Gamma irradiation was conducted for just over 41 days, with total absorbed gamma dose ranging from ~45 MGy for the lightest salts to ~170 MGy for ThF₄ and UF₄. By measuring the gas pressure within salt-filled capsules during irradiation, it was possible to quantify radiolytic gas production for all salt samples except UF₄. Production rates are reported as the salt G-values, measured as number of fluorine molecules produced per 100 eV of energy absorbed (molecules F₂/100 eV). The G-values of the salts were found to be G(LiF) ~0.004, G(BeF₂) ~0.009, G(ThF₄) ~0.021, and G(FLiBe-UF₄) ~0.005.

Four fluoride fuel salt samples (78LiF-22ThF₄) in graphite crucibles were irradiated in the HFR Petten for a duration of 508 Full Power Days under the name SALIENT-01 (SALt Irradiation ExperimeNT). Goal of the experiment was to gain experience with the design of liquid salt experiments and the handling of the salts before and after irradiation. Specific research goals for SALIENT-01 are (i) to confirm claims of good fission product retention in the salt, (ii) to obtain size distributions for noble metal particles using Transmission Electron Microscopy and (iii) to assess possible interactions between fuel salt and fine-grained nuclear graphite, as well as possible uptake of fission products by the graphite. Here the design and irradiation history of the experiment are discussed together with plans for post-irradiation examinations. Limitations in representativeness of this experiment and capsule irradiations in general are discussed as well as follow-up actions to improve the quality of future irradiations.

NRG together with JRC Karlsruhe has set out to perform a series of molten fuel salt irradiations in the High Flux Reactor (HFR) Petten, to build up irradiation experience with molten salt samples, the handling of irradiated salt and the treatment of salt waste produced by these irradiations. As a first step, an irradiation of small 78LiF-22ThF₄ fuel samples in graphite crucibles is being conducted under the name SALIENT-01 (SALt Irradiation ExperimeNT). The specific goals for SALIENT-01 are (i) to confirm claims of good fission product retention in the salt by post-irradiation Knudsen-cell effusion by fission product release measurements, in particular with respect to Cs and I which dominate the radioactive release during thermal transient and fuel failure in conventional reactor fuel, as for example occurred in the Fukushima accident, (ii) to obtain size distributions for noble metal particles using Transmission Electron Microscopy and (iii) to assess possible interactions between fuel salt and fine-grained nuclear graphite, as well as possible uptake of fission products by the graphite. The SALIENT-01 irradiation has started in August 2017 and is projected to finish in August 2019. In this contribution the design, irradiation history and status of the experiment are treated, as well as plans for post-irradiation examinations. The limitations of the experiment will also be discussed, as well as the follow-up actions taken to assess its representativeness and to improve the quality of future molten salt irradiations.