Molten Fluoride Salts Research Articles

High-temperature corrosion of helium ion-irradiated Ni-based alloy in fluoride molten salt

A comparison of the effect of helium-ion-radiation damage on a Ni–Mo–Cr–Fe alloy exposed to high-temperature (750°C) corrosion in eutectic LiF–NaF–KF molten salt has been made. Microstructural examination showed that both the corroded-only and irradiated and corroded samples exhibit the characteristics of intergranular corrosion. However, helium ion irradiation introduces defects such as helium bubbles and cavities into the microstructure, greatly increasing intragranular corrosion of the irradiated sample. The thickness of the remaining corrosion zone for the irradiated sample was much greater than that for the unirradiated sample.

Estimation of Melting Points for Some Binary and Tertiary Fluoride Molten Salts

Mixtures of fluoride molten salts such as LiF + BeF2 (Flibe) and LiF + NaF + KF (Flinak) have been proposed as tritium breeders for fusion reactors and heat-transfer fluids for high-temperature fission reactors. The melting point of mixed molten salts is important in fusion and fission reactor designs. An analytical method using the activity coefficient of the Margules’ equation and another method using an equilibrium constant when a new phase appears in the phase diagram are proposed for calculating melting points according to whether or not the new phase appears. First, the melting points of pure fluorides of LiF, NaF, KF, and BeF2 are investigated in detail, and uncertainties in the thermodynamic properties of the targeted molten salt mixtures are clarified. Then, the melting points of some binary- and tertiary-component fluoride molten salt mixtures of LiF + NaF (Flina), LiF + KF (Flik), NaF + KF (Fnak), LiF + NaF + KF (Flinak), LiF + BeF2 (Flibe), and NaF + BeF2 (Fnabe) are analytically investigated to enhance their wider application in fusion and fission reactors. Estimated melting points are compared with experimental data reported previously. Estimation errors are within 3.0 K (0.3%) for the pure fluorides and within 34 K (5.2%) for the binary or tertiary fluoride mixtures. Although estimation errors for the Flinak system are larger than those of previous reference data, the present estimation does not include an accommodation factor, and the parameter values included in the estimation are consistent with other thermodynamic data. The values of the activity coefficient used for estimation of the Flinak system and the equilibrium constant included in estimation of the Flibe and Fnabe systems are consistent with relevant thermodynamic properties. Therefore, the present method can be applied to estimate melting points for a range of multicomponent fluoride mixtures.

Electrochemical Study of Corrosion in High Temperature Molten Salts

Concentrating solar power systems are projected to need to produce energy with a cost of $0.06 / kWh when implemented on a very large scale. It is generally believed that temperatures of the heat transfer fluids will need to be higher than 800oC to drive an advanced Brayton cycle or Rankine cycle with superheated steam to reach this target. However, current state-of knowledge of corrosion science indicates a lack of published data on the impact of molten salts at temperatures above 800°C on corrosion of molten salt contacting materials such as steels or high temperature alloys and ceramics. This work investigates the corrosion mechanism of selected alloys and ceramics in chloride and fluoride molten salts used for high operating temperature heat transfer fluids using a combination of electrochemical techniques, theoretical calculations, and thermodynamic predictions. The electrochemical characterizations included linear polarization resistance scans, electrochemical impedance spectroscopy, and cyclic potentiodynamic polarization. Results show that the Fe-Ni-Cr alloys tested have high general corrosion and require corrosion inhibition. Advanced materials, such as SiC-SiC and TZM, have been shown to have much lower corrosion and good passivation behavior. Experimental tests using coupon immersion in the molten salts support the mechanism determined from electrochemical testing that the main element being oxidized is Cr. Analysis of samples also indicates selective Cr oxidation creates porous layers in the sample. The electrochemical impedance tests suggest that the corrosion is diffusion controlled.

Electrochemical Fluorination of Uranium to UF6

The development of efficient and environmentally benign methods to reprocess used nuclear fuel will be enabling technologies for the nuclear renaissance. The primary large scale method for processing used nuclear fuel is aqueous and generates large volumes of high level radioactive waste. The use of electrochemical and fluorination methods for reprocessing used nuclear fuel have been proposed and researched by various groups, but combinations of electrochemical and fluorination methods have not been investigated. Fluorinating agents are extremely potent oxidizers and may electrochemically fluorinate metallic used nuclear fuel with no input of electricity. A galvano-fluorination process has been developed at Savannah River National Laboratory. This electrochemical non-aqueous process enables efficient gas phase uranium recovery while allowing for control of the reaction. Thermodynamic modeling of the electrochemical fluorination process has shown UF6 to be the first gaseous fission product produced during fluorination. Further thermodynamic modeling was used to determine the feasibility of oxide reduction required for oxide fuel before electrochemical fluorination and for experiments. A chemical reactor system for electrochemical fluorination of U with NF3 in a fluoride molten salt was constructed along with vessels for product collection and analysis. This work has proven that an electrochemical reaction happens between U and NF3 which confirms the feasibility of electrochemical fluorination and the development of a processing scheme.

Use of Molten Salt Fluxes and Cathodic Protection for Preventing the Oxidation of Titanium at Elevated Temperatures

The current study demonstrates that it is possible to protect both solid and liquid titanium and titanium alloys from attack from air by cathodically polarizing the titanium component using an electro-active high-temperature molten salt flux and a moderate polarization potential. The electrolytic cell used comprises a cathode of either solid titanium or liquid titanium alloy, an electrolyte based on molten calcium chloride or fluoride salt, and an anode consisting of an inert oxygen-evolving material such as iridium metal. The new approach renders possible the processing of titanium at elevated temperatures in the presence of oxygen-containing atmospheres.

Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

Production and release of ISOL beams from molten fluoride salt targets

In the framework of the Beta Beams project, a molten fluoride target has been proposed for the production of the required 101318Ne/s. The production and extraction of such rates are predicted to be possible on a circulating molten salt with 160MeV proton beams at close to 1MW power. As a most important step to validate the concept, a prototype has been designed and investigated at CERN-ISOLDE using a static target unit. The target material consisted of a binary fluoride system, NaF:LiF (39:61mol.%), with melting point at 649°C. The production of Ne beams has been monitored as a function of the target temperature and proton beam intensity. The prototype development and the results of the first online tests with 1.4GeV proton beam are presented in this paper.

Failure analysis of 316L stainless steel crucible by molten fluoride salt interaction with clay bonded silicon carbide

Detailed investigation of a recently failed static corrosion test involving molten eutectic LiF–NaF–KF salt at 850°C contained in 316L stainless steel crucibles shows that a single sealed crucible leaked molten salt into the electric furnace due to mechanical failure, causing a chemical reaction between the molten salt and the clay bonded silicon carbide insulation used in the furnace. As a result, corrosive vapors of Na2SiF6, K2SiF6, SiF3, and F2, were formed. These vapors reacted with the stainless steel crucibles and lead to the formation of a porous corrosion crust and the eventual catastrophic structural failure of all nine sealed 316L stainless steel crucibles.

First-principles molecular dynamics modeling of the molten fluoride salt with Cr solute

Fluoride salts and their interactions with metals are of wide interest for the nuclear community. In this work, first-principles molecular dynamics (FPMD) was employed to study both pure molten fluoride salt and fluoride salt with dissolved solute Cr ions (a common corrosion product) at high temperature (823–1423K). Two types of molten fluoride salts, namely flibe (LiF–BeF2) and flinak (LiF–NaF–KF), with the Cr0, Cr2+ and Cr3+ ions were chosen as a target system for the FPMD modeling. The prediction of thermo-kinetic properties of pure fluoride salt, such as the equilibrium volume, density, bulk modulus, coefficient of thermal expansion, and self-diffusion coefficient, provide useful extensions of existing data and verify the accuracy of the FPMD simulation in modeling of fluoride salts. The FPMD modeling of solute Cr in fluoride salt shows the effect of Cr valence on diffusivity and local structure in the salt.

Reheat Air-Brayton Combined Cycle Power Conversion Off-Nominal and Transient Performance

Because molten fluoride salts can deliver heat at temperatures above 600 °C, they can be used to couple nuclear and concentrating solar power heat sources to reheat air combined cycles (RACC). With the open-air configuration used in RACC power conversion, the ability to also inject natural gas or other fuel to boost power at times of high demand provides the electric grid with contingency and flexible capacity while also increasing revenues for the operator. This combination provides several distinct benefits over conventional stand-alone nuclear power plants and natural gas combined cycle and peaking plants. A companion paper discusses the necessary modifications and issues for coupling an external heat source to a conventional gas turbine and provides two baseline designs (derived from the GE 7FB and Alstom GT24). This paper discusses off-nominal operation, transient response, and start-up and shutdown using the GE 7FB gas turbine as the reference design.

Improvement of stacking order in graphite by molten fluoride salt infiltration

The structure of molten fluoride salt infiltrated graphite has been studied by X-ray diffraction and Raman spectroscopy. The full width at half maximum and asymmetry degree of the (002) diffraction peak decrease with the increase of salt weight gain in the infiltration indicates the improvement of the stacking order of the graphite by the molten salt infiltration. The shape of G′ resonance of Raman spectra further indicated the enhancement of the stacking order of graphite by molten salt infiltration. The shift of the (002) diffraction peak position to the higher 2θ angle and the increase of the coefficient of thermal expansion reveal that the improvement of the stacking order is resulted from the compression of the graphite matrix by the infiltrated salt network. The compressive stress induced by the salt network in the graphite matrix could reduce the d-spacing fluctuation between graphite basal planes and close the Mrozowski crack generated by the anisotropy of graphite crystal, thus increase the crystallite size along c-axis.

Solution of heat removal from nuclear reactors by natural convection

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He+ ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He+ ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He+ ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy.

Heat transfer and pressure drop characteristics of molten fluoride salt in circular pipe

FLiNaK salt, eutectic mixture of fluorides of lithium, sodium and potassium, is a potential candidate coolant for high temperature reactors (HTRs). In this paper, a CFD analysis has been performed to study the heat transfer and pressure drop characteristics of FLiNaK salt flowing in a circular pipe for various regimes of flow. Simulation is performed with the help of in-house developed CFD code, NAFA. The calculated local Nusselt number (Nu) and friction factor along the pipe have been compared with the available correlations. The dependence of Nusselt number on Reynolds number (Re) has also been predicted and compared. Pressure drop along the pipe has been compared with various correlations. This study shows that the considered correlations for heat transfer and pressure drop are suitable to predict the heat transfer and pressure drop behavior of FLiNaK salt.

Zirconium Effect on the Corrosion Behavior of 316L Stainless Steel Alloy and Hastelloy-N Superalloy in Molten Fluoride Salt

A corrosion test was performed on 316L stainless steel alloy (316L) and Hastelloy-N superalloy (Hastelloy-N) at 850°C for 1000 h in static molten fluoride salt, 46.5LiF-11.5NaF-42KF (mol %) with Zr additions. The interactions between the graphite sample and the tested alloys in the molten salt were also analyzed. The results show that Zr addition to the salt caused the deposition of a pure Zr coating on 316L and Hastelloy-N. The formation of this coating was followed by interdiffusion between the Zr deposit and the substrates. A thicker Zr deposit was observed on Hastelloy-N samples compared to 316L due to the larger electromotive potential difference between Ni/Zr than that between Fe/Zr. The interdiffusion subsequent to Zr deposition led to the formation of a coating composed of a Ni/Zr intermetallic phase even on the iron-based 316L. This intermetallic coating on the two alloys acted as a barrier layer for Fe and Cr outward diffusion. Zr3NiO and ZrO2 phases were also observed on the coating surfaces and in the coatings, respectively. The graphite sample, on the other hand, had no direct and significant effect on the corrosion behavior of the alloys and the coating formation on the alloys.

Investigation of the MSFR core physics and fuel cycle characteristics

The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.

Sealing nuclear graphite with pyrolytic carbon

Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830°C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR).

Abundant thorium as an alternative nuclear fuel

It has long been known that thorium-232 is a fertile radioactive material that can produce energy in nuclear reactors for conversion to electricity. Thorium-232 is well suited to a variety of reactor types including molten fluoride salt designs, heavy water CANDU configurations, and helium-cooled TRISO-fueled systems. Among contentious commercial nuclear power issues are the questions of what to do with long-lived radioactive waste and how to minimize weapon proliferation dangers. The substitution of thorium for uranium as fuel in nuclear reactors has significant potential for minimizing both problems. Thorium is three times more abundant in nature than uranium. Whereas uranium has to be imported, there is enough thorium in the United States alone to provide adequate grid power for many centuries. A well-designed thorium reactor could produce electricity less expensively than a next-generation coal-fired plant or a current-generation uranium-fueled nuclear reactor. Importantly, thorium reactors produce substantially less long-lived radioactive waste than uranium reactors. Thorium-fueled reactors with molten salt configurations and very high temperature thorium-based TRISO-fueled reactors are both recommended for priority Generation IV funding in the 2030 time frame. • Thorium is an abundant nuclear fuel that is well suited to three advanced reactor configurations. • Important thorium reactor configurations include molten salt, CANDU, and TRISO systems. • Thorium has important nuclear waste disposal advantages relative to pressurized water reactors. • Thorium as a nuclear fuel has important advantages relative to weapon non-proliferation.

Investigation on corrosion behavior of Ni-based alloys in molten fluoride salt using synchrotron radiation techniques

Ni-based alloys have been selected as the structural materials in molten-salt reactors due to their high corrosion resistance and excellent mechanical properties. In this paper, the corrosion behavior of some Ni-based superalloys including Inconel 600, Hastelloy X and Hastelloy C-276 were investigated in molten fluoride salts at 750 °C. Morphology and microstructure of corroded samples were analyzed using scanning electron microscope (SEM), synchrotron radiation X-ray microbeam fluorescence (μ-XRF) and synchrotron radiation X-ray diffraction (SR-XRD) techniques. Results from μ-XRF and SR-XRD show that the main depleted alloying element of Ni-based alloys in molten fluoride salt is Cr. In addition, the results indicate that Mo can enhance the corrosion resistance in molten FLiNaK salts. Among the above three Ni-based alloys, Hastelloy C-276 exhibits the best corrosion resistance in molten fluoride salts 750 °C. Higher-content Mo and lower-content Cr in Hastelloy C-276 alloy were responsible for the better anti-corrosive performance, compared to the other two alloys.

Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF–KF–ZrF4 molten fluorides

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr–1Nb–1Sn–0.1Fe) were used as anode materials and electrorefining was performed in a LiF–KF eutectic 6wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.