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DOE OSTI2021

Quantify Sodium Fluoride / Beryllium Fluoride Salt Properties for a Liquid Fueled Fluoride Molten Salt Reactor (CRADA C2018-18168 Final Report)

ThorCon is developing a thermal thorium-uranium fueled molten salt reactor that uses a fuel salt consisting of NaF - BeF 2 - ThF 4 - UF 4 - UF 3 , which uses 19.75% enriched uranium. It is not a breeder reactor but requires regular additions of fissile material. A thorough understanding of the thermophysical properties of the fuel salt at the beginning (BOL) and end (EOL) of its life is critical to establishing its thermal hydraulic behavior and enables ThorCon to design a system with the required natural convection. Experimentally determining the heat capacity and thermal conductivity of the fuel salt will provide data necessary for detailed system design and safety analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Measurement of the Kinematic Viscosity of Melted Mixtures of Sodium, Lithium, and Beryllium Fluorides and the Effect of the Eutectic Additive Cerium Trifluoride on the Viscosity

Experimental data on the kinematic viscosity and liquidus temperature of certain fused mixtures of sodium, lithium, and beryllium fluorides were obtained, by the method of damping of torsional oscillations of a cylinder with the experimental melt, in validation of the concept of molten-salt incinerator reactors for incinerating transuranium elements from spent nuclear fuel from light-water reactors. The measurements were conducted from the liquidus temperature to 800°C. The effect of the addition of cerium trifluoride on the viscosity of the eutectic melt 58NaF–15LiF–27BeF{sub 2} was studied. It was shown that the additive appreciably reduces the viscosity at low temperature and also lowers the liquidus temperature. The parameters characterizing the temperature dependence of viscosity were determined for the experimental melts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2019

Optimizing the 8 Li yield for the IsoDAR Neutrino Experiment

The focus of this paper is on optimizing the electron-antineutrino source for the IsoDAR (Isotope Decay at Rest) experimental program. IsoDAR will perform sensitive short-baseline neutrino oscillation and electroweak measurements, among other Beyond Standard Model searches, in combination with KamLAND and/or other suitable detectors. IsoDAR will rely on the high-Q β– decay of the 8 Li isotope for producing electron-antineutrinos, created mainly via neutron capture in an isotopically enriched 7 Li sleeve surrounding the Be target. In particular, this paper examines the performance, defined in terms of absolute 8 Li (or, equivalently, electron-antineutrino) production rate, of various candidate sleeve materials, including a lithium-fluoride, beryllium-fluoride mixture (“FLiBe”) sleeve and a homogeneous mixture of lithium and beryllium (“Li-Be”). Furthermore, these studies show that the 8 Li yield can be increased substantially by employing a Li-Be sleeve and therefore motivate significant changes to the nominal IsoDAR design.

47 OTHER INSTRUMENTATION↗

DOE OSTI2024

ICP-MS For Analysis of Lithium Isotopic Ratios in Materials Highly Enriched in 7 Li (CRADA 626) Final Report

The intent of this project was to determine appropriate methodologies for analysis of lithium isotopes for 1) quasi-online monitoring of isotope separations, and 2) high-precision characterization of end-products, including lithium fluoride and lithium-beryllium fluoride (FLiBe). Traditionally, isotopic analysis requires the use of expensive, sophisticated instrumentation with often relatively long analytical timelines. We determined that the use of cheaper, smaller, and easier-to-operate instruments are sufficient for rapid analysis of lithium isotopes to monitor isotope separation efficiencies. In addition, we found that the analytical precision afforded by these lower-cost instruments is comparable to the more sophisticated (i.e., more expensive) instrumentation when the isotopic composition of lithium becomes highly depleted in 6Li. This finding will ultimately help lower costs of manufacturing of depleted lithium that is required for operation of certain designs of Generation IV nuclear reactors.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2019

Materials Data on Be2Ni(NF7)2 by Materials Project

(BeF2)2Ni(NF5)2 crystallizes in the monoclinic P2_1/c space group. The structure is zero-dimensional and consists of four beryllium fluoride molecules and two Ni(NF5)2 clusters. In each Ni(NF5)2 cluster, Ni2+ is bonded in an octahedral geometry to six F1- atoms. There are four shorter (1.76 Å) and two longer (2.26 Å) Ni–F bond lengths. N4+ is bonded in a trigonal non-coplanar geometry to three F1- atoms. There is two shorter (1.39 Å) and one longer (1.50 Å) N–F bond length. There are five inequivalent F1- sites. In the first F1- site, F1- is bonded in a single-bond geometry to one N4+ atom. In the second F1- site, F1- is bonded in a single-bond geometry to one Ni2+ atom. In the third F1- site, F1- is bonded in a single-bond geometry to one N4+ atom. In the fourth F1- site, F1- is bonded in a distorted bent 120 degrees geometry to one Ni2+ and one N4+ atom. In the fifth F1- site, F1- is bonded in a single-bond geometry to one Ni2+ atom.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Materials Data on Be(NF2)2 by Materials Project

BeF2(NF)2 crystallizes in the orthorhombic Pnma space group. The structure is zero-dimensional and consists of four beryllium fluoride molecules and eight monofluoroamine molecules.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Magnetic Bearing Pumps for Molten Salt Fusion Energy Devices

The Commonwealth Fusion Systems (CFS) ARC reactor will employ a novel liquid immersion blanket design that uses a molten lithium salt as both the tritium-breeding material and the vacuum vessel coolant. fluoride lithium beryllium (FLiBe) (a lithium fluoride–beryllium difluoride mixture) is currently the leading salt candidate. Its low-Z components provide good moderation, it has adequate heat transfer properties and low electrical conductivity, and it has good neutron multiplication properties, which further enhance tritium production.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Corrosion characteristics of monolithic SiC materials in beryllium-bearing molten fluoride salt

In this work, chemical vapor deposited SiC single crystal and polycrystals were exposed to a static beryllium-bearing molten fluoride salt, 2LiF–BeF 2 for 500 and 1000 h at 750 °C. Overall, the single crystal and polycrystalline SiC materials experienced slight weight loss and nominal surface etching regardless of the material type. Localized pitting and nonuniform Si removal from surface regions were associated with trace metal impurities and oxygen. Cross-sectional observation revealed minimal subsurface changes in the microstructure. The corrosion damage processes were similar to those reported for molten LiF–KF–NaF salt with trace impurities, indicating that the impurities significantly affect SiC corrosion.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Corrosion characteristics of silicon carbide fiber-reinforced composites in beryllium-bearing molten fluoride salt

Corrosion of SiC-fiber reinforced SiC matrix composites in 2LiF-BeF 2 molten salt was examined following static molten salt exposure for 1000 h at 650 and 750 °C. Composites were fabricated with either Tyranno SA3 or Hi-Nicalon type-S fibers and a chemical vapor infiltration SiC matrix. Both composites showed minimal weight loss after salt exposure and maintained shape. Corrosion was characterized by comprehensive electron microscopy and Raman spectroscopy. In conclusion, based on the results of experimental and thermodynamic analysis, a corrosion mechanism for SiC/SiC is proposed where both the salt and SiC materials play a key role in the phase stability of the composite.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Fluoride salt static freezing study

Molten salt reactor designs such as the FHR must consider the safety implications of freezing, requiring the development of thermal hydraulics models that account for freezing behavior. Pipe models that couple heat transfer, fluid dynamics, and phase change physics will require verification and validation. A static freezing study is being conducted to provide an initial set of validation data with LiF-BeF{sub 2} (FLiBe). The study will also be performed with LiF-NaF-KF (FLiNaK) in order to quantify the distortion that arises from using FLiNaK as a non-toxic surrogate salt for thermal-hydraulic studies. In order to develop the numerical model, accurate inputs for the thermal hydraulic properties of the two salts of interest need to be measured near the freezing point. For this reason, a series of three experiments including Differential Scanning Calorimetry (DSC) measurements, the Density Measurement Experiment, and the Static Freezing Experiment are being conducted to determine these thermo-physical properties at temperatures near the freezing point. The Static Freezing Experiment is also designed to determine the degree of supercooling that FLiBe and FLiNaK experience. Supercooling is a temperature depression below the freezing point that the liquid experiences before freezing initiates. Increased supercooling in a liquid flowing through a pipe leads to increased dendritic freezing and an increased likelihood of pipe blockage and freeze-shut. Based on previous research on supercooling in water and non-fluoride salts, supercooling is likely affected most strongly by cooling rate and sample size (as well as salt purity). No data has been generated previously on supercooling in fluoride salts such as FLiBe and FLiNaK. The results of this set of DSC, Density Measurement and Static Freezing experiments will be used in an ongoing fluoride salt freezing study being conducted to develop pipe freezing models for molten salts to be used in safety analysis. The thermal hydraulic properties measured in this study will serve as inputs for all future models. With the supercooling behavior of both salts quantified, it can be assessed whether supercooling must be included in future thermal hydraulic models. The validated Static Freezing Experiment model can then serve as the framework for a pipe freezing model and eventually a 1D system code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Neutronic Tests of Fluoride Salt Based MSR/FHR Coolants

Research on molten salt reactor (MSR) technology based on fluoride salts as a convenient way of higher actinides incineration and/or utilization of thorium fuel has been performed at UJV Rez and later Research Centre Rez in Czech Republic since 1999. Necessary steps towards to the deployment and possible commercialization were examined in past investigations. Neutronic testing is an inseparable part of the development and was at first carried out in a series of experiments with LiF-NaF salt (60 mol% LiF, 40 mol% NaF) subjected to a neutron field shaped by different moderators within the framework of the EROS experiments. Later, neutron spectrum was measured in this salt and significant variations in calculation to experiment C/E comparison were discovered. Further tests were performed with Flibe (66 mol% LiF, 33 mol% BeF{sub 2}) reactor coolant salt containing enriched lithium (99.994 mol% Li-7), which was provided by Oak Ridge National Laboratory (ORNL). This Flibe salt originally comes from the Molten Salt Reactor Experiment (MSRE) and was supplied to the Czech Republic within the framework of a collaborative agreement on MSR and fluoride-salt-cooled high-temperature reactor (FHR) research and development (R and D) concluded between the Czech Ministry of Industry and Trade and US Department of Energy in 2012. All experiments in this work were carried out at the LR-0 reactor at Research Centre Rez. The LR-0 is a pool-type zero power light water reactor with large and flexible experimental capabilities due to its design and fuel construction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

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–BeF{sub 2}–PuF{sub 3}, LiF–BeF{sub 2}–UF{sub 4}, and LiF–BeF{sub 2}–ThF{sub 4}–UF{sub 4}. 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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Separation of Actinides from Lanthanides in Molten Fluorides by the Modulated-Current Electrolysis

Recently, the Molten Salt Reactor (MSR) is taking a renewed attraction as one of the reactor systems highlighted by the Generation IV International Forum. It is a non-classical (liquid fuel) system which can be operated as a breeder within the {sup 232}Th-{sup 233}U fuel cycle with the very interesting breeding factor. The necessary precondition for its proper functionality is the online reprocessing within the MSR's fuel cycle. A lot of effort is given to the development of the on-line pyrochemical reprocessing technology. Among the processes considered to be suitable for this process, the electrochemical separations are highlighted. Within this context, it is necessary to study electrochemical behavior of both fissile material and fission products in the molten salt systems and to develop the processes of their extraction from the system at various stages of the whole cycle. In this contribution, the study of the electrochemical separation of U from Gd in the LiF-CaF{sub 2} molten system by the modulated current electrolysis is presented. The study of MSR-related electrochemistry will continue as a part of the new Czech MSR R and D program. The knowledge of the electrochemical behavior of actinides and lanthanides in molten fluorides is necessary to the development of any electrochemical separation process. In our previous work, electrochemical characteristics of uranium, thorium, as well as that of several lanthanides, were measured in LiF-BeF{sub 2} (the basic melt considered for the MSR concept) and Be-free melts (LiF-NaFKF, LiF-NaF, LiF-CaF{sub 2}), and added to available literature data. If the fluoride carrier melt is electrochemically stable enough, it is possible to deposit the element of interest in its basic, metallic state (e.g. uranium in all types of fluoride melts tested) by the electrolysis. In the case of insufficient electrochemical stability of the fluoride melt in which the actinides and lanthanides are dissolved for the experiments, it is possible to use a reactive working electrode made of a material which creates alloys with the element to be deposited. An alloying reaction can shift the deposition potential in a favorable direction and therefore achieve the separation of an element from the system. The use of reactive electrode can also improve the adhesion of the deposit. When two or more electroactive species are present in the system at the same time, the efficiency of their separation by an electrolysis depends on how different the deposition potentials of these two species are. If we input certain modulation to the electrolytic current, we can improve overall result of an electrolytic experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Comparing Sensitivity/Uncertainty Analysis Results for LR-0 Salt Experiments with Salt Reactor Models

Recent reactor physics measurements conducted at Research Centre Rez (RC Rez) used the LR-0 low power nuclear reactor to investigate neutronic impacts and uncertainties of FLiNa (60 mol% LiF and 40 mol% NaF) and FLiBe (2 LiF + BeF{sub 2}) salt targets. This included using FLiBe originally from the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL), which contained enriched lithium (99.994 mol% {sup 7}Li). In collaboration with RC Rez, ORNL performed sensitivity/uncertainty (S/U) analyses of these LR-0 experiments with salt targets. Molten salt reactor technologies can be categorized into two general types: solid-fueled or liquid-fueled. Solid-fueled molten salt technologies are perhaps best represented by fluoride-salt-cooled high-temperature reactors (FHRs), which are solid-fueled and salt-cooled. They are potentially promising because they offer high-temperature outlet heat using a low-pressure, high-power-density system (relative to high temperature gas-cooled reactors). However, FHRs have never been built; therefore, no direct neutronic benchmarks exist for code or cross section validation work. Liquid-fueled molten salt reactors (MSRs) use a liquid salt as both fuel and coolant. MSRs can be designed to use a variety of fertile or fissile species as nuclear fuel, different neutron energy spectra, and a multitude of approaches for adding or removing material from the salt. The MSRE generated some operational data for fluoride salt MSRs, but no formal benchmarks currently exist based on the MSRE data, and the data available do not cover all operational conditions. Given these data limitations, ORNL performed S/U analyses using application models for FHR and MSR concepts to (1) identify potential sources of bias in FHR and MSR simulations resulting from cross section uncertainties and to (2) compare the sensitivity of neutron multiplication in FHR and MSR concepts to cross section data on an energy-dependent basis for specific nuclides against similar sensitivity results from LR-0 with a FLiBe salt target. In the future, these S/U analyses could inform the design of additional salt experiments or provide part of the basis for follow-up work to refine nuclear data. Please see the final project report for additional details beyond this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Dynamic Analysis of the Next Generation Molten-Salt Breeder Reactor System

The conceptual design of a 1000 MWe two-fluid molten-salt breeder reactor (MSBR) station was performed during 1966-1970 at Oak Ridge National Laboratory (ORNL) with the goal of using the U-233 fuel cycle. This modular design with four reactor modules has an electrical output of 250 MW/module. The abundance of Th-232 was attractive in converting it to U-233 in a thermal breeder reactor. The two-fluid reactor is graphite-moderated 'with a {sup 7}LiF-BeF{sub 2}-UF{sub 4} fuel salt circulated through the core and a {sup 7}LiF-ThF{sub 4}-BeF{sub 2} blanket salt circulated through separate flow channels distributed throughout the core, as well as in a surrounding under-moderated region'. This reactor design has the advantages of not producing Pu isotopes and easy chemical separation of U-233 in the blanket material. The low fuel cost and the graphite replacement cost are attractive for such a design, with an estimated cost of $141/kW at 1968-dollar value. Potential non-proliferation features of MSBR, the low costs of raw materials, and established chemical separation of fission products are important factors in further development of MSBRs as an alternative to current light water reactors. These reactors are often referred to as liquid fluoride thorium reactor (LFTR). The current research on the modeling of the MSBR is based on an earlier work on the preliminary dynamic model of the MSBR concept. Revisions are made to the original model. The physics parameters have been modified for the current modeling effort based on additional information including more recent values of model parameters from and several typographical misprints in equations and values. Details of this effort will be presented in a full follow-up paper. This work presents the detailed nodal models of reactor fluid core, blanket fluid loop, graphite moderator, fuel fluid heat exchanger, and the blanket fluid heat exchanger. A set of simulations has been performed to validate the model using the commercial software MATLAB{sup TM}-Simulink. Simulations were performed for the cases of source perturbation and reactivity perturbation. The model is inherently nonlinear. A description of the MSBR conceptual design, along with a tabulation of relevant system parameters, is presented. The results of simulation demonstrate stable behavior of the MSBR dynamics, temperature feedback effects, and load-following capability of the MSBR system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs

The SCALE code system was used to model, deplete, and compare several different TRISO-fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a Fluoride-Lithium-Beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were build and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burn-up dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17x17 fuel. Overall, the results showed that short term decays are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burn-up are of only secondary importance. Although this analysis only focuses on activities in becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long- term time periods post-irradiation shows that the reactor type and maximum burn-up have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in post- release scenarios; normal fuel handling operations; and spent fuel transport, storage and disposal. Of particular interest, the results in this report show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long term results will also be incorrect, especially for areas such as permanent disposal. Even though this report focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Computational Fluid Dynamics Analysis of the Molten Salt Tritium Transport Experiment Test Section

Tritium, a radionuclide produced through neutron capture by lithium and other elements (beryllium and fluoride) in molten salts, presents unique challenges to radionuclide release. This is true for both fusion energy breeder blankets and molten salt fission reactors. The fundamental understanding of tritium transport is crucial to the safe design and operation of these reactors. The Molten Salt Tritium Transport Experiment (MSTTE), currently under construction at Idaho National Laboratory, aims to investigate tritium transport phenomena using a forced-convection fluoride salt loop. This loop is designed to study various transport mechanisms, such as permeation through metals and gas-liquid interactions, and is intended to support future research on tritium extraction units. A critical aspect of the MSTTE loop design is ensuring a fully developed velocity profile before the fluid reaches the permeation test section where measurements are made. This study employs computational fluid dynamics to model the salt flow behavior within the MSTTE permeation test section. A realizable k-ε turbulent model with enhanced wall treatment is used to simulate the single-phase, vertical upward flow of molten salt FLiNaK under isothermal conditions. The simulation results indicated flow distortion and underdeveloped profiles at all planned flow rates within the test section due to the 85-deg sharp bend. To address this issue, a reduced diameter with a reducer and expander and a flow conditioner are investigated to achieve fully developed flow. The analysis showed that the flow conditioner successfully corrected the flow profile, achieving fully developed behavior at a flow rate of 50 liters per minute (LPM). In conclusion, this research enhances our understanding of flow dynamics in molten salt systems and contributes to optimizing tritium transport control technologies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

A laboratory-scale process for producing dilithium beryllium tetrafluoride (FLiBe) with dissolved uranium tetrafluoride

Flibe Energy, Incorporated (FEI)'s conceptual Lithium Fluoride Thorium Reactor (LFTR) incorporates a chemical processing facility aimed at recovering uranium and other valuable volatile radionuclides while managing harmful radionuclides from the used fuel. The fuel utilized in this reactor is a combination of dilithium beryllium tetrafluoride (Li 2 BeF 4 or FLiBe) and uranium tetrafluoride (UF 4 ), (FLiBe/U). FEI's plan involves extracting the uranium and other valuable volatile fluoride-forming radionuclides using nitrogen trifluoride (NF 3 ). To facilitate laboratory-scale testing of uranium extraction using NF 3 and address the toxicity and physical hazards associated with beryllium and beryllium fluoride (BeF 2 ), we used a two-step process to prepare the simulated fuel salt. The first step entailed thermally decomposing ammonium beryllium tetrafluoride [(NH 4 ) 2 BeF 4 ] (ABeF) through a nominal 3-step process, combined with appropriate amounts of lithium fluoride (LiF) and UF 4 , resulting in the formation of beryllium fluoride (BeF 2 ). In the second step, the mixture was repeatedly melted and frozen at the melting point of FLiBe to prepare the eutectic FLiBe with dissolved UF 4 . Although the concept appears straightforward, the production of FLiBe/U involved various challenges. These challenges included transporting the gaseous decomposition products of ABeF, hydrogen fluoride (HF) and ammonia (NH 3 ), while preventing the formation of ammonium fluoride (NH 4 F). Additionally, it was necessary to control the reaction between the higher-than-anticipated water content in the commercial ABeF with NH 3 , HF, and the condensed NH 4 F, protect UF 4 from forming an unknown black compound, select suitable structural materials to mitigate fluoride corrosion, address the risks associated with beryllium toxicity through equipment design and operational protocols, and monitor process conditions. This article provides an account of the thermal decomposition chemistry observed in the commercial ABeF, describes the FLiBe/U production apparatus, describes the experiences and process refinements developed to prepare FLiBe/U, and presents our characterizations of prepared FLiBe/U.

Ammonium beryllium fluoride thermal decomposition↗