LiF BeF2 Research Articles

Numerical analysis of the viscosity of ZrF4–BeF2 molten salts for nuclear transmutation using fusion neutron sources

This study analyzed the influence of ZrF4 on the viscosity of a ZrF4–BeF2 binary salt towards developing BeF2-based molten salts for nuclear transmutation targets. Specifically, we performed molecular dynamics simulation using a polarization ion model and evaluated the viscosity, static structure, and applicability of the Stokes–Einstein equation for ZrF4–BeF2 compared with LiF–BeF2. Our study revealed that ZrF4–BeF2 exhibits higher viscosity than LiF–BeF2 because each ionic bond's longer lifetime extends the structural relaxation time. Furthermore, it is suggested that the ZrF and BeF relaxations contribute to the viscosity mechanism of ZrF4–BeF2. Most of the elements in long-lived fission products (LLFPs) are multivalent ions such as Zr, Pd, and Sn, which bind strongly to anions and are not expected to contribute much to reducing the viscosity of BeF2 mixed salts. In contrast, Cs, also an LLFP, is a monovalent ion, and CsF can reduce the viscosity of these salts. Mixing particular LLFPs with CsF will be one of the principles in designing BeF2-based salts for transmutation.

Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor

Abstract FLIBE (LiF–BeF2) is both a nuclear reactor coolant and a solvent for fertile or fissile materials. FLIBE can also dissolve a variety of fissile and fertile materials such as uranium, thorium and plutonium and has the ability to serve as liquid fuel for molten salt reactors (MSRs). It’s very high thermal capacity and chemical stability are among its other valuable properties. In addition, the low atomic weights of lithium, beryllium and to a lesser extent fluorine make FLIBE an effective neutron moderator. In this study, cross-section values were determined by using various level density models (constant temperature + Fermi gas, back-shifted Fermi gas, generalized superfluid, microscopic level density models) for reactions of 19F(n, α)16N, 19F(n, p)19O, 35Cl(n, α)32P, 35Cl(n, p)35S and 79Br(n, α)76As, 79Br(n, p)79Se, 127I(n, α)124Sb, 127I(n, p)127Te using TALYS 1.95 code and these datas are compared with the values in the EXFOR database.

Fuel Cycle of the LiF–BeF2 Molten Salt Actinide Burner Reactor

Fuel Cycle of the LiF–BeF2 Molten Salt Actinide Burner Reactor

A study of plutonium fluoride solubility in LiF–BeF2 melt for the substantiation of a molten salt reactor for Np, Am, Cm transmutation

A study of plutonium fluoride solubility in LiF–BeF2 melt for the substantiation of a molten salt reactor for Np, Am, Cm transmutation

Thermophysical Properties of Several Molten Mixtures of the LiF–BeF2–UF4 System

Thermophysical Properties of Several Molten Mixtures of the LiF–BeF2–UF4 System

Comparative Analysis of Transmutation in a Burner Reactor Based on the Salts LiF–NaF–KF and LiF–BeF2

Comparative Analysis of Transmutation in a Burner Reactor Based on the Salts LiF–NaF–KF and LiF–BeF2

Thermal Properties of Li2BeF4 near Melting Point

The LiF–BeF2 system is used as a heat transfer medium in molten salt reactors (MSRs). The thermal diffusivity of Li2BeF4 was studied using the laser flash analysis (LFA) method in solid and transition states. While the thermal diffusivity is shown to decrease slightly in solid-state Li2BeF4, it drops significantly at temperatures close to phase transition. The heat capacity of Li2BeF4 was measured by differential scanning calorimetry (DSC). Some differences were observed between the results obtained in cooling and heating modes. Thermal conductivity was calculated using thermal diffusivity-, density-, and heat-capacity data. The good thermal conductivity of the Li2BeF4 compound in solid and liquid states justifies its use as a heat transfer medium for molten salt reactors.

The Viscosity of Molten Salts Based on the LiF–BeF2 System

The Viscosity of Molten Salts Based on the LiF–BeF2 System

Viscosity of molten salts based on the LiF–BeF2 system

Rotational viscometry with the FRS 1600 (Anton Paar, Austria) high-temperature rheometer was used to obtain temperature dependences of the dynamic viscosity of molten lithium and beryllium fluoride salts considered as candidate fuel and coolant compositions for the molten salt reactor (MSR) for burning long-lived actinides from the spent nuclear fuel of the PWR 1000/1200 pressurized water reactor. 0.66LiF–0.34BeF2 and (0.73LiF–0.27BeF2)+UF4 molten salt mixtures containing 1 and 2 mol.% UF4 were investigated with regard to the MSR intermediate and fuel circuits. Salt mixtures were prepared by the direct melting of components and certified using X-ray phase and elemental analysis. The «shear rate» parameter was selected according to the viscosity curves obtained in the studied melts at 700 °C. It was found that the viscosity does not depend on the shear rate in the range of γ = 6÷20 s–1. When measuring the temperature dependence of viscosity, the shear rate was 11 s–1. Viscosity values of LiF–BeF2–UF4 melts obtained from experiments in the temperature range from liquidus to 800 °C are described by the linear equation logη = a + b/t, but their temperature coefficients differ evidently, which indicates a significant dependence of the viscosity of these melts on composition and temperature. Viscosity values obtained for the 0.66LiF–0.34BeF2 melt agree with the available literature data within 7–10 % in the temperature range of 650–750 °C. With an increase in the LiF content, melt viscosity decreases: it is lower by 20 % in the 0.73LiF–0.27BeF2 melt at t = 650 °C. However, when 2 mol.% UF4 is added, the 0.73LiF–0.27BeF2+UF4 fuel salt viscosity increases by 10 % at the same temperature.

Corrosion behavior of GH3535 alloy in molten LiF–BeF2 salt

The corrosion behavior of GH3535 alloy in molten LiF–BeF2 salt was investigated in a thermal convection loop. Results indicate that NiF2 in molten LiF–BeF2 salt and O2 from leakage drive the corrosion of alloy. The corrosion of alloy at the hot section is severer than that at the cold one. The BeCr2O4 was detected on the surface of alloy at the cold section while not at the hot one. Reactions between Be and metallic fluorides result in the concentration of impurities in molten LiF–BeF2 salt significantly reduced, and no obvious corrosion occurred in the GH3535 alloy at different temperature.

First-principles molecular dynamics study of the behavior of tritium in molten LiF-BeF2 eutectic

Molten LiF–BeF2 (FLiBe) eutectic (67 mol% LiF and 33 mol% BeF2) plays the fundamental role in the molten salt reactors (MSRs). However, the residual 6Li nuclides capture neutrons and generate the radioactive tritium nuclides, thus the tritium management is essential for the safe operation of MSRs. In this work, first-principles molecular dynamics (FPMD) simulations are applied to the investigation of the behavior of tritium in molten FLiBe. It is proved that the calculated temperature dependence of density and the coordination structure of molten FLiBe agree well with the available experimental information. Thus, it is confident that FPMD can be applied to the investigation of behavior of tritium in molten FLiBe. Finally, the coordination and electronic structures, as well as transport characteristics of various tritium moieties are studied for the understanding of the behavior of tritium in molten FLiBe.

Inhibition effect of ZrF4 on UO2 precipitation in the LiF-BeF2 molten salt.

The dissolution–precipitation behavior of zirconium dioxide (ZrO2) in molten lithium fluoride–beryllium fluoride (LiF–BeF2, (2 : 1 mol, FLiBe)) eutectic salt at 873 K was studied. The results of the dissolution experiment showed that the saturated solubility of ZrO2 in the FLiBe melt was 3.84 × 10−3 mol kg−1 with equilibrium time of 6 h, and its corresponding apparent solubility product (K′sp) was 3.40 × 10−5 mol3 kg−3. The interaction between Zr(iv) and O2− was studied by titrating lithium oxide (Li2O) into the FLiBe melt containing zirconium tetrafluoride (ZrF4), and the concentration of residual Zr(iv) in the melt gradually decreased due to precipitate formation. The precipitate corresponded to ZrO2, as confirmed by the stoichiometric ratio and X-ray diffraction analysis. The K′sp was 3.54 × 10−5 mol3 kg−3, which was highly consistent with that from the dissolution experiment. The obtained K′sp of ZrO2 was in the same order of magnitude as that of uranium dioxide (UO2), indicating that a considerable amount of ZrF4 could inhibit the UO2 formation when oxide contamination occurred in the melt containing ZrF4 and uranium tetrafluoride (UF4). Further oxide titration in the LiF–BeF2–ZrF4 (5 mol%)–UF4 (1.2 mol%) system showed that ZrO2 was formed first with O2− addition less than 1 mol kg−1, and the precipitation of UO2 began only after the O2− addition reached 1 mol kg−1 and the precipitation of ZrO2 decreased the ZrF4 concentration to 0.72 mol kg−1 (3 mol%). Lastly, UO2 and ZrO2 coprecipitated with further O2− addition of more than 1 mol kg−1. The preferential formation of ZrO2 effectively avoided the combination of UF4 and O2−. This study provides a solution for the control of UO2 precipitation in molten salt reactors.

Solubility and precipitation investigations of UO2 in LiF–BeF2 molten salt

The solubility of UO2 in molten LiF–BeF2 (2:1 mol) (FLiBe) eutectic salt at 600 °C was studied by chemical and electrochemical methods. The results of dissolution experiment showed that the saturated solubility of UO2 in this melt was 2.37 × 10−3 mol/kg and its corresponding apparent solubility product (Ksp’) was approximately 1.67 × 10−5 mol3/kg3. When more Li2O were added to the FLiBe–UF4 system, the cathodic peak current of U(IV) in the electrochemical cyclic voltammetry (CV) curve accordingly decreased because of precipitate formation. The precipitate corresponded to UO2 as determined by stoichiometric ratio (concentration variation of U4+ and O2−) and X-ray diffraction (XRD) analysis. Compared to the chemical analysis method, the CV technique was confirmed to be more feasible for accurate determination of concentration of U(IV). Meanwhile, the Ksp’ value was also obtained to be 1.33 × 10−5 mol3/kg3 during the whole oxide titration procedure, which was highly consistent with that from the dissolution experiment. With the value of Ksp’, the allowable amount of dissolved oxide ions (oxide tolerance) can be theoretically estimated in the FLiBe–UF4 system.

Investigation of the local structure of molten ThF4-LiF and ThF4-LiF-BeF2 mixtures by high-temperature X-ray absorption spectroscopy and molecular-dynamics simulation.

The microscopic structures of ThF4-LiF and ThF4-LiF-BeF2 molten salts have been systematically investigated by in situ high-temperature X-ray absorption fine-structure (XAFS) spectroscopy combined with molecular-dynamics (MD) simulations. The results reveal that the local structure of thorium ions was much more disordered in the molten state of the ThF4-LiF-BeF2 salt than that in ThF4-LiF, implying that the Th and F ions were exchanged more frequently in the presence of Be ions. The structures of medium-range-ordered coordination shells (such as Th-F2nd and Th-Th) have been emphasized by experimental and theoretical XAFS analysis, and they play a significant role in transport properties. Using MD simulations, the bonding properties in the molten ThF4-LiF and ThF4-LiF-BeF2 mixtures were evaluated, confirming the above conclusion. This research is, to the best of our knowledge, the first systematic study on the ThF4-LiF-BeF2 molten salt via quantitative in situ XAFS analysis and MD simulations.

Np, Am, Cm Transmutation in Different Types of Reactors

Different approaches to the transmutation of Np, Am, and Cm in molten-salt reactors are compared: based on the eutectics LiF–NaF–KF (BZhSR), LiF–NaF–BeF2 (MOSART), and the salt LiF–BeF2 (ZhSR-S). In addition the efficiency of transmutation in a BREST-1200 type fast reactor is evaluated. It is shown that the highest possible transmutation efficiency can be realized in a reactor based on the salt LiF–NaF–KF (~306 kg/yr per 1 GW(t)). The results show that further study of the physical and chemical properties of LiF–NaF–KF is warranted, specifically, a study of the solubility of a fuel composition with fission products and an investigation of the possibility of developing a reactor with a fast neutron spectrum on the basis of this salt.

Influence of α-Al2O3 and AlF3 on pyrohydrolysis of Li3AlF6

In this study, Li3AlF6 was employed to simulate the molten salt LiF–BeF2 to explore its pyrohydrolysis behavior and that of its components, i.e., LiF and AlF3, respectively. The influence of the accelerators α-Al2O3 and AlF3 on the pyrohydrolysis of LiF and Li3AlF6 was investigated. Finally, the solid pyrohydrolytic products were characterized by means of X-ray diffraction, and the corresponding reaction mechanisms were proposed. These experimental results indicated that AlF3 was completely hydrolyzed to the corresponding oxide α-Al2O3 at 650 °C in 1 h, whereas the complete hydrolysis of LiF and Li3AlF6 required the assistance of either α-Al2O3 or AlF3 under the same conditions. The influence of the accelerator α-Al2O3 and AlF3 on the pyrohydrolytic behavior of Li3AlF6 provides references for future research studies on the pyrohydrolysis of LiF–BeF2 and multi-component molten salts.

Corrosion Resistance and Mechanical Stability of Nickel Alloys in Molten-Salt Nuclear Reactors

Choosing structural materials for molten-salt nuclear reactors is a high-priority problem. The results making it possible to choose the basic structural material for a molten-salt nuclear reactor are colligated in the present article. Forced and natural circulation setups operating under reactor and laboratory conditions with molten metal-fluoride salts with different compositions have been developed at the National Research Center Kurchatov Institute: LiF–NaF–BeF2–PuF3, LiF–BeF2–UF4, and LiF–BeF2–ThF4–UF4. Corrosion tests were conducted with domestically produced corrosion-resistant steel (12Kh18N10T, EP-164) and specially developed nickel-based alloys (KhN80M-VI, KhN80MTYu, KhN80MTV, et al.) as well as the alloy Hastelloy (USA), MONICR (Czechoslovakia) and EM-721 (France) in the working temperature range 600–800°C and mechanical loads up to 80 MPa on the samples.

Molten Salt Reactor for 99Mo Production

99Mo production in a molten salt reactor with fuel salt based on metal fl uorides (LiF–BeF2–UF4) is examined. The proposed method of production is based on a particularity of the behavior of gaseous fission products in salt melts. Molybdenum, together with some other metals, does not form stable compounds in fluoride melt. At least 50% of fragment molybdenum in a molten salt nuclear-physical facility will be in a gas-aerosol phase above the surface of the molten salt. Neutron-physical and thermohydraulic calculations were performed taking into account the specific features of a molten salt reactor. The optimal configuration of the reactor, the materials used, the fuel salt composition, and the fuel cycle are described.

First-principle investigation of the structure and vibrational spectra of the local structures in LiF–BeF2 Molten Salts

In this article, the identified local structures are investigated using Density Functional theory to reveal the effects of the interaction on the geometry and vibrational dynamics of the ternary-system. Ab initio molecular dynamics simulation is used to simulate the identified clusters in gas phase and the salts in condensed phase. Vibrational spectra are computed either through normal mode analysis or Fourier transform of the autocorrelation function of the total dipole moment. We find that two of the most intense bands can be taken as the signature of formation of the local structures (BeF42−, Be2F73−, and Be3F104−). The signature bands are found in the correspondent region in the spectra of the salts, indicating the existence of the local structures. Vibrational density of states spectra found that the feature frequency of Li atoms is as low as 400cm−1, suggesting it is not tightly bound to other species. Current study sheds some light into understanding of the interaction and dynamics of the molten salt.

Molecular dynamics studies of the structure of pure molten ThF4 and ThF4–LiF–BeF2 melts

The local structure around Th4+ ions in molten ThF4, and its mixtures with alkali and alkaline earth fluorides (LiF and LiF–BeF2) was investigated by molecular dynamics (MD) simulations with a polarizable force field. We found that the network formed determined to be almost exclusively composed of corner-sharing, F− linked [ThF8] in pure ThF4. The network became sparser when LiF was added to ThF4. In the ThF4–LiF–BeF2 system, both Be2+ and Th4+ ions formed a discrete coordination complex. The Be2+ ion is surrounded by four equidistant F anions and there are approximately 8 fluoride anions around Th4+. The Be–F coordination complex was more stable than that of Th–F. By adding Be2+ into the ThF4–LiF mixture, the thorium compound network can be made more compact. We also observed a higher coordination number and a smaller barrier in the ThF4–LiF–BeF2 melt. In addition, we calculated a smaller characteristic time from the cage-out correlation function than that calculated for the ZrF4–LiF–BeF2 system. We show that, in comparison with the ZrF4–LiF–BeF2 melt, the molten salt of ThF4–LiF–BeF2 has a loosely bound shell coordination structure.