Liquid Uranium Research Articles

Corrosion of iron in liquid uranium hexafluoride at 80 °C. Part I: Normal and abnormal experimental kinetics

The results reported here are the first on pure iron corrosion in liquid UF6 at 80 °C. Two kinetic behaviours have been observed: one led to micrometric scales (from micron to hundreds of microns after several months) and the other one to hundreds of nanometers for several months. The higher corrosion kinetics resulted in the presence of impurities such as NOxF complexes formed by interaction between the medium and the reactor material. These NOxF catalysed the corrosion reaction leading to a corrosion mechanism controlled by the cathodic reaction rate. Effect of impurities in UF6 coming from experimental conditions or nature of uranium ore should then be systematically and carefully checked. Whatever the corrosion kinetics and the presence of impurities, the nature of the layer was identical: a duplex fluoride scale composed of an iron rich layer, FeF2, and a uranium rich layer evolving over time from U2F9 to UF5.

Numerical Study of Porous Cylindrical Containment for the Fuel Element of a Centrifugal Nuclear Thermal Rocket

For deep-space propulsion and interplanetary exploration, the centrifugal nuclear thermal rocket (CNTR) has the ability to achieve a very high specific impulse (Isp) metric beyond that of conventional chemical rockets or solid-core nuclear thermal propulsion systems. The high Isp allows the rocket to use less propellant or achieve a higher velocity for shorter transit times. However, the cylindrical containment structure of a CNTR fuel element encounters extreme conditions, as it houses molten uranium at temperatures exceeding 1408 K, leading to challenges such as dissolution, chemical reactions, and thermal stresses that conventional materials struggle to withstand. This study aims to address this issue by analyzing appropriate materials for constructing the cylindrical containment component. The operating conditions of the annular porous medium that confines the liquid uranium in the centrifugal fuel element are simulated by conducting a comprehensive one-dimensional numerical analysis using a range of candidate porous materials, including Mo, W, zirconium carbide, and silicon carbide. The porous structure facilitates the flow of the hydrogen propellant into the internal molten uranium section, where it gains significant thermal energy while simultaneously cooling the cylinder. The containment cylinder has an internal temperature of 1478.1 K, exceeding the melting point of uranium, while the external gas temperature of the hydrogen propellant is much lower. This temperature difference induces significant thermal stresses in the cylinder. The porous containment cylinder made from molybdenum was able to maintain elastic deformation throughout the thickness of the cylinder, showcasing its ability to handle these extreme thermal stress conditions. Tungsten, on the other hand, experienced plastic deformation at the cylinder’s edges and elastic deformation through the middle radial locations. In contrast, the stresses experienced by the ceramic materials far exceeded their failure stress values, leading to brittle failure. These findings will help with the refinement of the CNTR design, edging it closer to practical implementation.

Interaction Between Uranium Trifluoride and the Mixture of Lithium and Beryllium Fluorides

Equilibrium potentials of uranium (U3+/U0) have been measured in the LiF-BeF2-UF3 melt as a function of temperature and uranium fluoride concentration. The empirical equation of uranium potential isotherms and polytherms have been obtained. Cathode polarization of uranium in the molten mixture of beryllium and lithium fluorides has been measured using the current switch off method form the stationary state. Uranium ions were found to have primarily the-valence of three in the LiF-BeF2 electrolyte in the studied temperature and uranium fluoride concentration ranges. A conditional standard potential of uranium (U3+/U0) in the molten mixture of lithium and beryllium fluorides was calculated relative to the reference fluorine electrode according to the experimentally obtained data on the equilibrium potentials of the uranium electrode in the LiF-BeF2-UF3 melt. Standard conditional changes in the Gibbs energy, enthalpy and entropy at the formation of uranium trifluoride from the elements in the form of dilute solutions were determined. The enthalpy of mixing of liquid uranium fluoride and the 0.73LiF–0.27BeF2 melt was calculated.

One-Dimensional Steady-State Thermal Model of CNTP Reactor

A more advanced form of nuclear propulsion known as centrifugal nuclear thermal propulsion (CNTP) promises increased propellant temperatures that could lead to a high specific impulse in the range of 1500 to 1800 s with hydrogen. This design has the potential of opening opportunities to perform missions to destinations much farther than currently possible. However, the CNTP concept poses many engineering challenges due to the nuclear fuel operating at high temperature in a liquid phase. A one-dimensional, steady-state thermal model of the liquid uranium fuel has been constructed to understand the limitations of this concept and the potential design considerations. Three related basic designs are considered, and key design parameters are varied in order to predict the temperature levels and void fractions across the liquid uranium pool.

Steady-state analyses of an smDFR with coupled Serpent-OpenFOAM calculation

The small modular dual fluid reactor (smDFR) adopts a mixture of liquid uranium and plutonium chloride as fuel, with lead as a coolant. With its smaller size, lower weight, and inherent safety, the smDFR can be used as a mobile energy supply for vehicles or ships. Analyses of the smDFR are still in the early stage based on a single or separated physical field, which can hardly capture the impact of the flowing fuel on the neutron field. The coupled neutronic-fluid dynamic calculation is needed for reliable multidimensional, multi-field distribution of the steady-state smDFR core property. In this work, the Monte Carlo code Serpent 2 and the CFD code OpenFOAM are coupled via a Python interface, whereby the neutron field information can be exchanged with the fluid field in a region-average manner. The correctness and reliability of the coupled code are verified with two cases in the pressurized water reactor background, and the results show good agreement. The coupled code is applied to the steady-state analyses of an smDFR, a single-channel case, and a 1/6-core case. The results show that the temperature distribution of the fuel and coolant caused a lower drift of the axial power peak compared to the uncoupled cosine shape power distribution. The 1/6-core results show that the peak temperature of the fuel reaches 980 K at the top-center of the reactor core. Moreover, the influence of fuel thermal properties and inlet velocity are studied with rated power. Through the comparison of mapping schemes, the results can identify the dependence of specific parameters and help to reduce mapping errors. This work shows that through a coupled neutronic-CFD code, the multidimensional and multi-physical fields of the smDFR can be better represented, providing important references for the future design and analyses of smDFR systems.

Some comments on Herndon’s nuclear georeactor

Abstract Light was thrown on the nuclear aspects of Herndon’s georeactor. His thesis is that a power producing droplet of liquid uranium, possessing a radius of about 5 km, operated during the last 4.5 billions of years in the center of the Earth and producing 3–6 TW of thermal power. It could be shown that indeed initial criticality was possible due to the high U-235/U-238 enrichment of approximately 30% at that time. Furthermore, the U-238/Pu-239/U-235 conversion cycle guarantees also a stabilized U-235/U-238 enrichment of about 10% during the whole history of Earth. For this, a constant critical neutron flux in the realm of 108 n/(cm2s) has to be present to convert U-238 via the Pu-239 route and its eddecay into U-235 with a conversion ratio of exactly unity. Herndon’s georeactor is compared with other exotic fission reactors in Nature, too, but the decisive answer whether such a georeactor really exists needs further – particularly chemical, thermodynamical and geological – research work.

(Invited) Electrochemical Recycling of Used Nuclear Fuel

Current nuclear energy systems produce approximately twenty percent of the electricity generated in the U.S. and account for approximately fifty-five percent of the carbon-free electricity. As the U.S. strives for even lower carbon emissions such as net-zero carbon emission by 2050, nuclear energy will play an increased role in providing baseload electricity to the nation. An opportunity to implement advanced reactor and fuel cycle technologies that offer enhanced resource utilization, waste management and non-proliferation features will be available as the role of nuclear energy expands to meet the growing demand of carbon-free electricity. For example, the U.S. can transition from the current once-through fuel cycle to a closed fuel cycle with used fuel recycling that provides actinides for energy production and encapsulates fission product waste in durable, engineered waste forms.Argonne National Laboratory has a rich-history in developing new methods to recycle used nuclear fuels that dates back to the late 1950s. This early work focused on developing methods to recycle the actinides extracted from used nuclear fuel and closing the nuclear fuel cycle. Early techniques consisted of melting the used metallic fuel discharged from the Experimental Breeder Reactor II in lime-stabilized zirconia crucibles to produce a liquid uranium alloy that could be separated from the resulting oxide dross containing the fission product elements. The resulting uranium alloy was recycled to produce fresh nuclear fuel. Subsequently, more elegant techniques such as molten salt – liquid metal extraction processes were developed to recover and recycle the actinides present in the oxide dross, and to separate the actinides from the fission product elements present in used metallic fuel. In both cases, the actinides were recycled in fresh nuclear fuel and the fission products were sequestered in waste forms.Argonne’s research and development efforts in used fuel recycling expanded to include areas such as oxide to metal conversion via chemical reduction in a molten salt flux and developing new processes for recycling the actinides extracted from nitride and carbide fuels. Work in this area evolved to include the design and development of electrochemical technologies for recovering actinide metals from used nuclear fuels and an electrolytic reduction process for the conversion of metal oxides to metallic forms.The transition of used nuclear fuel processing from chemical to electrochemical methods established the need for fundamental electrochemical data for the actinide and fission product elements in molten salt media, provided opportunities for the design and demonstration of purpose-built electrochemical processing equipment, and allowed for the use of process monitoring and control technologies to enable efficient electrochemical cell operations. Actinide metal nucleation and growth in a molten LiCl - KCl eutectic salt is an example of the type of fundamental data that has been collected to elucidate the behavior of actinides during the electrodeposition process. Results from those experiments revealed that actinide nucleation transitions from progressive to instantaneous growth with increasing actinide chloride concentration in the molten salt. This behavior influenced the design of electrochemical processing equipment. Other data such as actinide ion diffusion coefficients were also determined from the experiments.Electroanalytical process monitoring and control techniques are being developed for several molten salt applications. These techniques are ideally suited for molten salt applications due to the specificity of chemistry information that is produced, simplicity of equipment design, and functionality at elevated temperature and in radiation environments. Argonne developed a multi-function sensor and methodology that yields reproducible electrochemical data such as the redox potential of the salt and the identities of species present in the system for applications in used fuel processing and monitoring heat transfer fluids.Argonne’s molten salt activities span the range from fundamental property measurement to pilot-scale process demonstrations. The presentation will provide a survey of Argonne’s electrochemical fuel recycling activities including actinide electrochemistry, process design and development, and electroanalytical sensor development.ACKNOWLEDGMENTS: The submitted abstract has been created by UChicago Argonne, LLC, Operator of Argonne National Laboratory (“Argonne”). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02-06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.

In situ liquid SIMS analysis of uranium oxide

Trace analysis of nuclear materials in solid particles collected in the environment or particles in liquid slurry generated in nuclear material manufacturing processes can pinpoint elemental, organic, and isotopic signatures of nuclear fuel cycle activities and processes. Such information can support nuclear safeguards programs by increasing our ability to detect undeclared nuclear materials, routine activities for safeguarding at declared facilities, and illicit activities. However, trace radioactive material analysis in liquids and slurries is challenging using bulk approaches. For example, one drawback of sensitive analysis such as inductively coupled plasma mass spectrometry (ICP‐MS) is that sample is consumed or destroyed as a result of the technical approach. We developed a vacuum compatible microfluidic interface to enable surface analysis of liquids and solid–liquid interactions using time‐of‐flight secondary ion mass spectrometry (ToF‐SIMS). In this work, we illustrate the initial results from the analysis of liquid uranium oxide standard solutions using in situ liquid SIMS. Because the liquid SIMS analysis is almost nondestructive, the same sample can then be analyzed by other analytical techniques or saved for future reference. Consequently, multimodal analysis is possible. Our results demonstrate that in situ liquid SIMS can be used as a new approach to analyze radioactive materials in liquid and slurry forms of relevance to diverse applications.

First principles study of liquid uranium at temperatures up to 2050 K

Uranium compounds are used as fissile materials in nuclear reactors. In present day reactors the most used nuclear fuel is uranium dioxide, but in generation-IV reactors other compounds are also being considered, such as uranium carbide and uranium mononitride. Upon possible accidents where the coolant would not circulate or be lost the core of the reactor would reach very high temperatures, and therefore it is essential to understand the behaviour of the nuclear fuel under such conditions for proper risk assessment. We consider here molten metallic uranium at several temperatures ranging from 1455 to 2050 K. Even though metallic uranium is not a candidate for nuclear fuel it could nevertheless be produced due to the thermochemical instability of uranium nitride at high temperatures. We use first principles techniques to analyse the behaviour of this system and obtain basic structural and dynamic properties, as well as some thermodynamic and transport properties, including atomic diffusion and viscosity.

Sustainable supply of 99Mo source in a 2 MW molten salt reactor using low-enriched uranium

The 99Mo production in a 2 MW molten salt reactor using liquid low-enriched uranium (LEU) fuel has been evaluated. The batch-wise extraction period of 99Mo is optimized to be one day corresponding to 9415 6-day Ci/week of the 99Mo production rate. The required amount of uranium is only 4.77 kg annually. The required chemically reprocessed amount of FPs is about 58.4 g annually, accounting for only 4.9% of the solid LEU target method under the identical production capacity of 99Mo.

3D EVKLID/V2 Code Aided Simulation of Severe Accidents

When fuel rods melt during a severe accident, the movement of the structural and fuel materials along fuel assemblies can be spatially non-uniform, so that a three-dimensional thermohydraulic model is implemented in the EVKLID/V2 code as part of the HYDRA-IBRAE/LM module. The transport of the same sets of components and their mixtures as in the one-dimensional version of the HYDRA-IBRAE/LM module can be calculated in the model: liquid sodium, sodium vapor or vapor-gas mixture of sodium vapor and non-condensable gases, solid lead, liquid lead, solid uranium dioxide, liquid uranium dioxide, hard stainless steel, liquid stainless steel, steel vapor. The three-dimensional model is implemented in a cylindrical coordinate system, which makes it easier to include the geometric dimensions and parameters of a particular fuel assembly (number and diameter of fuel rods, lattice pitch, and others) and the core. A description is given of the basic system of equations describing the motion of the components of the destroyed core in the three-dimensional r–z–φ geometry, and its numerical realization. Examples of test calculations showing the serviceability of the model are presented.

Evaluation of Experimental Studies in the U-Rich Region of the U-Pu Phase Diagram

Upon the re-examination of two experimental studies, additional thermophysical data pertaining to the uranium-rich portion of the uranium-plutonium binary phase diagram have been elucidated. The first study involved the analysis of solidification cooling curves following the processing of U-Pu alloys. Two phase transitions (liquidus and bcc) were evaluated and the results agreed very well with the established boundaries. The second study was actually performed over sixty years ago but still warrants recognition due to the paucity of thermodynamic data on the U-Pu system to date. From the investigation, the activity coefficients of plutonium in liquid uranium from 1813 K to 2042 K were derived assuming Henry’s law from which the partial molar excess Gibbs energy change of plutonium was obtained: \( \Delta \bar{G}^{\text{ex}} = R*T*\ln \gamma^{\text{o}}_{\text{Pu}} = - \;76694 + 39.2*T\left( {{\text{J}}/{\text{mol}}} \right). \)

Phase field model of uranium carbide solidification through a combined KKS and orientation field approach

A Phase Field model is developed combining the Orientation Field approach to modeling solidification with the Kim, Kim, Suzuki method of modeling binary alloys. These combined methods produce a model capable of simulating randomly oriented second phase dendrites with discrete control of the solid–liquid interface energy and thickness. The example system of carbon in a liquid uranium (U) melt is used as a test for the model. The formation of uranium carbide within a liquid U melt is simulated for isothermal conditions and compares well with experiments.

Thermodynamic Properties of Condensed Uranium Dioxide

In the context of the development of the IVTANTHERMO information and reference system, heat capacity equations have been derived for solid and liquid stoichiometric uranium dioxide in the temperature ranges of 298.15–3130 K and 3130–8000 K, respectively, on the basis of an analysis of experimental data available in the literature. The appearance of new experimental data on the UO2 enthalpy and heat capacity up to 8000 K made it necessary to recalculate the temperature dependences of the heat capacity. The refined values of thermodynamic functions (heat capacities, entropy, enthalpy increments, and the reduced Gibbs energy) obtained by numerical integration of these equations in a range of 100–8000 K have been entered into the IVTANTHERMO database.

Thermodynamic properties of Pb3U11O36

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.

Carbon diffusion in molten uranium: an ab initio molecular dynamics study

In this work we used ab initio molecular dynamics within the framework of density functional theory and the projector-augmented wave method to study carbon diffusion in liquid uranium at temperatures above 1600 K. The electronic interactions of carbon and uranium were described using the local density approximation (LDA). The self-diffusion of uranium based on this approach is compared with literature computational and experimental results for liquid uranium. The temperature dependence of carbon and uranium diffusion in the melt was evaluated by fitting the resulting diffusion coefficients to an Arrhenius relationship. We found that the LDA calculated activation energy for carbon was nearly twice that of uranium: 0.55 ± 0.03 eV for carbon compared to 0.32 ± 0.04 eV for uranium. Structural analysis of the liquid uranium–carbon system is also discussed.

Magnetic Lyogels for Uranium Recovery from Wet Phosphoric Acid

The work introduces composite magnetic materials designed to capture uranium from wet-process phosphoric acid (WPA) containing 6 M H3PO4, 2% H2SO4, sodium fluoride, and metal salts of Fe(III) and Al(III). The materials include poly(vinyl chloride) (PVC) covalently modified with N,N-diethyldithiocarbamate (DEDTC) or O,O-diethyldithiophosphate (DEDTP) moieties by nucleophilic substitution of the >C–Cl bonds of PVC. To maintain the polymer processability, the maximum substitution degree was kept below 42%. The modified PVC formed stable organic gel (lyogel) materials with liquid uranium extractants such as di(2-ethylhexyl)phosphoric acid (DEHPA) or a liquid mixture of trialkylphosphine oxides, Cyanex 923. To impart the magnetic recoverability to the lyogels, iron nanoparticles (20–50 nm) coated by carbon for chemical stability were incorporated. The resulting magnetic lyogels contain variable contents of liquid extractants, maintain particle shape, exhibit very low leaching of the extractants, and are chemic...

Corrosion of iron in liquid uranium hexafluoride

ABSTRACTHere is reported a dedicated and reliable experimental setup which has been designed to study the corrosive effects of liquid uranium hexafluoride (UF6) on pure metallic iron at 80°C and 3 bars of pressure. Three corrosion tests have been carried out, ranging from several hundred to almost one thousand hours. No uniform corrosion was observed on early-stage samples. However, a nodule facies consisting in FeF2 and UFx (with 4 < x < 5) deposits is observed at the surface of the iron samples. These FeF2 nodules grow with time and merge, leading finally to the formation of an almost continuous FeF2 layer. The thickness of the layer over time seems to follow a parabolic kinetics law, finally reaching 15 µm after about one thousand hours of immersion in liquid UF6. After the nodules have merged, many cracks can be observed in the FeF2 layer, probably due to the growing stresses as the corrosion layer is formed.

Energy-free machine learning force field for aluminum

We used the machine learning technique of Li et al. (PRL 114, 2015) for molecular dynamics simulations. Atomic configurations were described by feature matrix based on internal vectors, and linear regression was used as a learning technique. We implemented this approach in the LAMMPS code. The method was applied to crystalline and liquid aluminum and uranium at different temperatures and densities, and showed the highest accuracy among different published potentials. Phonon density of states, entropy and melting temperature of aluminum were calculated using this machine learning potential. The results are in excellent agreement with experimental data and results of full ab initio calculations.

Investigation of α-phase and liquid uranium by the method of quantum molecular dynamics

The thermodynamic properties of α-phase and liquid uranium have been investigated by the method of quantum molecular dynamics. The isotherms and equilibrium properties (pressure, temperature, and internal energy) along isochores and isotherms have been calculated for densities of 16.5–41.1 g/cm3 and temperatures up to 60000 K. The dependence of the thermal pressure on temperature and internal energy is analyzed. The melting boundary of α-phase uranium is estimated. The shock adiabat is calculated for pressures up to 1.5 TPa based on the equation of state derived here. The results obtained are in good agreement with the published experimental data and the results of calculations according to different theoretical models.