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In order to progress in the development of Lead-cooled Fast Reactors, from the safety point of view it is essential to understand the chemical compatibility between liquid lead and uranium oxide. In the present work, entropy and heat capacity of Pb3U11O36, a possible ternary compound coming from fuel-coolant chemical interaction, were determined for the first time. Entropy at 298.15 K was obtained from low temperature heat capacity measurements using the Physical Property Measurement System (PPMS) in the temperature range 2–300 K, while the high temperature heat capacity has been measured by a drop calorimeter from 373 K to 1200 K. The experimental thermodynamic properties were compared with the values computed by means of DFT-GGA simulations, obtaining a very good agreement.

The low-temperature heat capacity of (U 1-y ,Am y )O 2− x solid solution with y = 0.0811 and 0.2005 and x = 0.01–0.03 was determined from a minimum of 12.52 K up to 297.1 K and from 9.77 K up to 302.3 K, respectively, using hybrid adiabatic relaxation calorimtry. The low temperature heat capacity results of the investigated system revealed the absence of the magnetic transition specific for UO 2 in the temperature region of 30 K. Since there are no experimental data available for AmO 2 in this temperature region, the results obtained for the intermediate compositions are validated based on the experimental data of UO 2 end-member and the low-temperature heat capacity computation of AmO 2 . In the measured temperature interval, excess heat capacity was observed for the two investigated intermediate compositions, which is concluded to be dominated by self-radiation effects at very low temperature.

PuF3 was synthetized by hydro-fluorination of PuO2 and subsequent reduction of the product by hydrogenation. The obtained PuF3 was analysed by X-Ray Diffraction (XRD) and found phase-pure. High purity was also confirmed by the melting point analysis using Differential Scanning Calorimetry (DSC). PuF3 was then used for thermodynamic assessment of the PuF3-LiF system. Phase equilibrium points and enthalpy of fusion of the eutectic composition were measured by DSC. XRD analyses of selected samples after DSC measurement confirm that after solidification from the liquid, the system returns to a mixture of LiF and PuF3.

The authors regret to inform that the thermodynamic data [Formula presented] of the pure compound CsI(l) was not reported correctly in Table 3. The corresponding enthalpy of formation should read: [Formula presented]

Mixed uranium and americium dioxides (U, Am)O 2±x are candidates as possible transmutation targets for generation IV reactors. In this work, the enthalpy increments of this solid solution were measured in the 470–1750 K temperature range by drop calorimetry for Am/(Am + U) ratios equal to 0.32, 0.39, 0.49, 0.58 and 0.68. Then, the heat capacity functions were obtained by derivation of the enthalpy data. The results of this work were compared to the heat capacity and enthalpy functions reported in the literature for the UO 2 [1] and AmO 2 [2] binary oxides and for the U 0.9 Am 0.1 O 2±x , U 0.8 Am 0.2 O 2±x mixed oxides [3]. From the obtained trend, it was found out that an excess contribution to the enthalpy increment appears for T > 1100 K in the compositions with Am/(Am + U)≥0.4 and a possible explanation attributing this effect to oxygen hypostoichiometry is provided. Finally, to verify the hypothesis, thermodynamic computations based on the CALPHAD method were performed for AmO 2-x under air and the results confirmed that the source of the excess contribution is the formation of oxygen vacancies.

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.

Investigations on the phase relation in the Pb-U-O ternary system have shown the formation of three lead uranium(VI) oxides: PbUO 4 , Pb 3 UO 6 , and Pb 3 U 11 O 36 . We present here the synthesis, thermal expansion, low- and high-temperature heat capacity, and thermal diffusivity of two representative phases (PbUO 4 and Pb 3 UO 6 uranates). Combining these results, a value for their thermal conductivity was derived and reported here for the first time.

Quantum chemical calculations of the molecular properties of the UO 2 (OH) 2 molecule are presented. From the results the thermodynamic properties of this gaseous species have been calculated, and these have been used to re-evaluate the existing literature study on the volatilization of U 3 O 8 in steam, allowing to derive the enthalpy of formation of the UO 2 (OH) 2 molecule.

The mixture LiF-ThF4-UF4-PuF3 (77.5–6.6-12.3–3.6 mol%), a fuel option for the Molten Salt Fast Reactor (MSFR), has been measured by differential scanning calorimetry (DSC) for determination of phase transitions, and by Knudsen effusion mass spectrometry (KEMS) for measurement of partial vapour pressures. The boiling point of the mixture was determined by extrapolation of total vapour pressure to 1 bar. Thermodynamic calculations were performed and compared with the experimental results. Novel experimental data on pure plutonium trifluoride are presented: melting point, vaporization enthalpy, vapour pressure and ionization energies by electron impact.

The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas–gas and gas–aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

This study presents an original approach of oxygen stoichiometry determination during high temperature (>2000 K) measurements of vapour pressure using the Knudsen effusion mass spectrometry technique. The method has been developed taking into account the vapour pressure measurements of series of (U1-x,Pux)O2-δU1-x,Pux)O2-δ samples with x(Pu) = 0.25, 0.5, 0.75 together with pure UO2-δUO2-δ and PuO2-δPuO2-δ end-members coupled with equilibrium calculations based on thermodynamic assessment of the U–Pu–O system. The presented method consists of two steps; in the first step the oxygen potential of the oxide phase is determined based on the measured partial vapour pressures of UO(g), UO2(g), PuO(g) and PuO2(g) gaseous species and during the second step the thus determined oxygen potential is linked with the matching oxygen stoichiometry of the sample. From the obtained results it has been demonstrated that it is possible to accurately estimate the oxygen stoichiometry of the mixed oxide fuel samples knowing the description of the oxygen potential of the corresponding end-members only.

A selected composition of the initial fuel of the Molten Salt Fast Reactor (MSFR) is assessed by differential scanning calorimetry (DSC) for melting point determination and by Knudsen effusion mass spectrometry (KEMS) for vaporization behaviour. Partial vapour pressures and thermodynamic activities of the MSFR fuel mixture are discussed indicating departures from ideal behaviour, and further interpreted by phase equilibria calculations. The boiling point of the mixture is obtained extrapolating vapour pressure experimental results. New results on the vaporization behaviour of pure uranium tetrafluoride are presented, together with the ionization potentials of UF4 by electron impact.

The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

Thermodynamic assessments of the Th–U and Th–Pu binary systems are performed in this study according to CALPHAD method and together with the data on the U–Pu system, which has been already published in the literature, the Th–U–Pu ternary system is determined. All available experimental data from the literature were used for the binary system assessments and based on them the Gibbs energy properties of the unknown phases have been optimized. For the excess Gibbs energy of the liquid phase the ionic liquid model has been used, whereas for solid solutions the Redlich–Kister description was selected. In case of the Th–U and Th–Pu binary systems a very good agreement between the assessment and the experimental equilibrium data has been achieved, with 80% of equilibrium data reproduced within ±20K.

Samples of irradiated MOX of 44 GWd/tHM mean burn-up were prepared by core drilling at three different radial positions of a fuel pellet. They were subsequently heated in a Knudsen effusion mass spectrometry system up to complete vaporization of the sample (~2600 K) and the release of fission gas (krypton and xenon) and helium was measured. Scanning electron microscopy was used in parallel to investigate the evolution of the microstructure of a sample heated in the same condition up to given key temperatures as determined from the gas release profiles. A clear difference for fission gas release and microstructure was observed as a function of the radial position of the samples and therefore of irradiation temperature. A good correlation between the microstructure evolution and the gas release could be established as a function of the temperature of irradiation and (laboratory) heating., JRC.E.3-Materials research

We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. Molecular dynamics (MD) codes have been used to estimate properties of the molten salt system that are important for ADSMS design but have never been measured experimentally.Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated., JRC.E.3-Materials research

A thermodynamic description of the LiF-CeF3-ThF4 system is made in this study using a two-sublattice model for the description of the solid solution and a quasi-chemical model based on quadruplet approximation for the liquid phase. New calorimetric experimental data of the binary LiF-CeF3, CeF3-ThF4 and ternary LiF-CeF3-ThF4 systems have been obtained in this work justifying the calculated phase diagrams. Using the obtained thermodynamic assessment the concentration of PuF3 in the LiF-ThF4 melt was estimated based on the similarities with CeF3 and the melting behaviour of the initial molten salt fast reactor fuel was discussed., JRC.E.3-Materials research

The enthalpy increment data for the (Th,U)O 2 and (U,Pu)O 2 solid solutions are reviewed and complemented with new experimental data (400–1773 K) and many-body potential model simulations. The results of the review show that from room temperature up to about 2000 K the enthalpy data are in agreement with the additivity rule (Neumann-Kopp) in the whole composition range. Above 2000 K the effect of Oxygen Frenkel Pair (OFP) formation leads to an excess enthalpy (heat capacity) that is modeled using the enthalpy and entropy of OFP formation from the end-members. A good agreement with existing experimental work is observed, and a reasonable agreement with the results of the many-body potential model, which indicate the presence of the diffuse Bredig (superionic) transition that is not found in the experimental enthalpy increment data.

The enthalpy increments of PuPO 4 have been measured using drop calorimetry in the temperature range from 530 K to 1386 K. The heat capacity was derived from the obtained data and compared with heat capacity data obtained directly from differential scanning calorimeter measured in this study from 400 K to 1400 K. The recommended heat capacity of PuPO 4 was determined based on both techniques as: C p 0 ( J K - 1 mol - 1 ) = 126.600 + 32.999 × 10 - 3 ( T / K ) - 1.9503 × 10 6 ( T / K ) - 2 .

The α – β phase transition of a zirconium alloy doped with 1 mol% of niobium (E110 alloy) is investigated using differential scanning calorimetry. The onset and endpoints of the transition are identified from the measured heat flow signal and from the integration of the observed peak the extent of the α – β phase change is calculated as a function of temperature. The experiment has been performed at different heating rates and a shift of the onset temperature with increasing heating rate was observed. From the dataset the equilibrium transition curve has been extrapolated and compared with other types of zirconium-based cladding materials.

The present study describes the full thermodynamic assessment of the LiF–NaF–BeF 2 –ThF 4 –UF 4 system which is one of the key systems considered for a molten salt reactor fuel. The work is an extension of the previously assessed LiF–NaF–ThF 4 –UF 4 system with addition of BeF 2 which is characterized by very low neutron capture cross section and a relatively low melting point. To extend the database the binary BeF 2 –ThF 4 and BeF 2 –UF 4 systems were optimized and the novel data were used for the thermodynamic assessment of BeF 2 containing ternary systems for which experimental data exist in the literature. The obtained database is used to optimize the molten salt reactor fuel composition and to assess its properties with the emphasis on the melting behaviour.

In previous studies a thermodynamic description of the LiF–NaF–KF–RbF–CsF–LaF 3 system was presented. In order to add PuF 3 to this system the assessments of LiF–PuF 3 , NaF–PuF 3 , KF–PuF 3 , RbF–PuF 3 , CsF–PuF 3 and LaF 3 –PuF 3 binary phase diagrams have been made. In case of the LiF–PuF 3 and NaF–PuF 3 the assessments have been based on known experimental data. The other binary systems have not been measured yet and the thermodynamic description has been made using the excess parameters from the previously assessed binaries containing LaF 3 , which is considered as a proxy compound for PuF 3 . The main aim of this study is to analyze potential compositions for a molten salt fast burner fuel.

A full thermodynamic description of the quaternary NaCl–MgCl 2 –UCl 3 –PuCl 3 system, a Molten Salt Fast Breeder Fuel, is presented. The binary phase diagrams have been assessed in this study and the data were used to extrapolate the higher order systems. To optimize the excess parameters of the liquid phase the modified quasi chemical model has been used, while for the solid solution the classical polynomial model has been applied. From the obtained results a possible fuel composition for the Molten Salt Reactor has been evaluated.

International audience; Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel.Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

The LiF–ThF4–PuF3–UF4 system is the reference salt mixture considered for the Molten Salt Fast Reactor (MSFR) concept started with PuF3. In order to obtain the complete thermodynamic description of this quaternary system, two binary systems (ThF4–PuF3 and UF4–PuF3) and two ternary systems (LiF–ThF4–PuF3 and LiF–UF4–PuF3) have been assessed for the first time. The similarities between CeF3/PuF3 and ThF4/UF4 compounds have been taken into account for the presented optimization as well as in the experimental measurements performed, which have confirmed the temperatures predicted by the model. Moreover, the experimental results and the thermodynamic database developed have been used to identify potential compositions for the MSFR fuel and to evaluate the influence of partial substitution of ThF4 by UF4 in the salt.