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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

The ARC pilot plant conceptual design study has been extended beyond its initial scope to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B 0 = 9.2 T) tokamak that is approximately the size of JET (R 0 = 3.3 m). Taking advantage of ARC’s novel design – demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket – this follow-on study has identified innovative and potentially robust power exhaust management solutions. The superconducting poloidal field coil set has been reconfigured to produce double-null plasma equilibria with a long-leg X-point target divertor geometry. Here this design choice is motivated by recent modeling which indicates that such configurations enhance power handling and may attain a passively-stable detachment front that stays in the divertor leg over a wide power exhaust window. A modified VV accommodates the divertor legs while retaining the original core plasma volume and TF magnet size. The molten salt FLiBe blanket adequately shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as an effective single-phase, low-pressure coolant for the divertor, VV, and breeding blanket. Advanced neutron transport calculations (MCNP) indicate a tritium breeding ratio of ∼1.08. The neutron damage rate (DPA/year) of the remote divertor targets is ∼3–30 times lower than that of the first wall. The entire VV (including divertor and first wall) can tolerate high damage rates since the demountable TF magnets allow the VV to be replaced every 1–2 years as a single unit, employing a vertical maintenance scheme. A tungsten swirl tube FLiBe coolant channel design, similar in geometry to that used by ITER, is considered for the divertor heat removal and shown capable of exhausting divertor heat flux levels of up to 12 MW/m 2 . Several novel, neutron tolerant diagnostics are explored for sensing power exhaust and for providing feedback control of divertor conditions over long time scales. These include measurement of Cherenkov radiation emitted in FLiBe to infer DT fusion reaction rate, measurement of divertor detachment front locations in the divertor legs with microwave interferometry, and monitoring “hotspots” on the divertor chamber walls via IR imaging through the FLiBe blanket.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Density Measurements of Molten LiF–BeF 2 and LiF–BeF 2 –LaF 3 Salt Mixtures by Neutron Radiography

The densities of eutectic (LiF) 2 –BeF 2 and mixtures of this salt (FLiBe) with LaF 3 were measured by dilatometry and by neutron attenuation from 673 K to 1,073 K. Because LaF 3 has a limited solubility in FLiBe, it was necessary to determine the amount of LaF 3 in solution before the density could be determined. The FLiBe density determination was favorably benchmarked against the literature data. A simple comparison was not available for the LaF 3 –FLiBe mixtures, so extrapolation of published data was necessary based on analysis using the Molten Salt Thermal Properties Database-Thermochemistry, or MSTDB-TC, developed by the US Department of Energy. Solubilities for LaF 3 in FLiBe ranged from 1 to 4 mol % over 673 to 1,073 K. The salt system was heated and cooled over 24 h to evaluate potential changes in composition and hysteresis during the measurement. Changes in the meniscus were observed, and these were included in the correction for density determinations. Salt surface tension may have led to supersaturation of LaF 3 in the salt because the solubility curve was nonlinear with respect to the inverse temperature, as would be expected for an ideal system. Surface tension measurements are currently underway to test this hypothesis.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Initial Neutronics Investigation of a Chlorine Salt-Based Breeder Blanket

Tritium breeding blankets within D-T-fueled fusion reactors contain lithium compounds and typically require neutron multiplier materials to achieve a tritium breeding ratio (TBR) consistent with self-sustaining operation. Liquid breeder blankets have some advantages over solid blankets, and previous blanket studies have investigated liquid metal as well as liquid salt–based blankets. Liquid salts have reduced magnetohydrodynamic effects as compared to liquid metals, but typically have a lower TBR. Recently, advanced fission reactor concepts have considered chloride-based salts in their design, and there is a significant amount of research work occurring to study these salts. Chloride salts have previously been considered for fusion reactors, but studies have typically found lower breeding ratios than for fluoride salts, such as 2(LiF)-BeF 2 (flibe) so they have not been further developed. In this work, we use a one-dimensional cylindrical radiation transport model of a conceptual tokamak fusion reactor to investigate the neutronics feasibility of using a chloride salt–based blanket that uses chlorine enriched in 37 Cl, which has both a low capture cross section and a substantial (n,2n) cross section. It is found that chloride salts (LiCl mixed with BeCl 2 and/or PbCl 2 ) can potentially achieve a ~3% to 5% higher TBR than fluoride molten salts, notably flibe, in the absence of a solid multiplier. Including a solid multiplier, however, does narrow this advantage, with TBRs estimated within ~1% of flibe with a 2-cm Be multiplier. Chloride salts can also reach lower melting points than flibe, potentially improving the scope for the use of reduced activation ferritic-martensitic steel as a structural material. There is substantial uncertainty in the calculations driven by limited thermochemical data for the Cl salts, plus cross-section uncertainties. The production of 36 Cl through 35 Cl(n,g) and 37 Cl(n,2n) has the potential to challenge the waste disposal rating of the blanket. Calculations indicate that, while this is not an immediate showstopper, this case depends upon the exact waste disposal rating criteria used for 36 Cl. Further work could reduce these uncertainties with improved thermochemical data, higher-fidelity modeling for downselected salts, and more refined waste disposal calculations and regulatory guidance. Lastly, it must be recognized that, as for all molten salts, corrosion and chemistry can present appreciable technical challenges that require further assessment in developing a practical blanket concept, and also that the enrichment of chlorine presents an additional technical and supply chain challenge.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Optimal Chloride Salt Mixture for a Fusion Blanket

Deuterium-tritium fusion reactors cannot operate for a significant period without a closed tritium fuel cycle, according to a recent National Academies of Sciences (NAS) report on bringing fusion reactors to the US electrical grid. This fact places breeder blankets as one of the foundational systems for self-sustained fusion reactor operation. The typical functional requirements for a breeder blanket system include producing tritium, absorbing kinetic energy, transporting thermal energy, and being environmentally attractive. State-of-the-art research on liquid blankets has converged to primarily focus on (LiF) 2 and BeF 2 (FLiBe) molten salts and a metallic eutectic of lead and lithium (PbLi), but “virtually all of the technologies related to the tritium fuel cycle are at a low technological readiness level”. This work sought to explore optimum blanket configurations as it aligned with the Oak Ridge National Laboratory (ORNL) FY 2023 Laboratory Directed Research and Development Program’s research priority of developing and expanding the current understanding of fusion blanket science and technology. This purpose of this work was to address ORNL research priorities and NAS recommendations by investigating novel liquid blanket materials that could provide self-sustaining operation and draw on experience from research on molten salts used for advanced fission reactors, concentrated solar, and thermal energy storage. The hypothesis when proposing this research was that there could be chloride-based blanket designs that can exceed the tritium breeding ratios of (FLiBe) molten salt blankets while reducing the use of Be (FLiBe), avoiding the generation of HF (FLiBe), and minimizing magnetohydrodynamic (MHD)-perturbed flow fields (PbLi). The fastest and most cost-effective path to deploying liquid fusion breeder blankets could be from maximizing the synergistic technological overlap between fusion, fission, concentrated solar, and thermal energy storage industries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Radiative Heat Transport and Optical Characterization of High Temperature Molten Salts

The goal of this project was to advance the field of molten fluoride radiative heat transfer (RHT), through optical property measurements, heat transfer loop construction and operation, and computational fluid dynamics simulations. The following outcomes have been accomplished. A comprehensive review has been conducted to compile all available RHT-relevant optical absorption data for molten fluorides. The various measurement methods have also been compiled and critiqued. Thermal emission measurement methods have been refined and documented. A unique thermal emissivity measurement facility has been constructed and used to measure the emissivity of flowing FLiBe-exposed SS316 samples. A slight increase has been observed in the emissivity compared to unexposed SS316. A dual reflectance-emission setup was partially demonstrated for the purpose of molten fluoride absorption measurement. Preliminary infrared edge data for FLiNaK was obtained. A visible near-infrared (NIR) measurement setup has been constructed and will be used to measure absorption of corrosion products in FLiBe. A new multi-band thermal radiation method using blackbody-weighted spectral bands for the absorption-coefficient was deployed to investigate the thermal performance of the UW Natural Circulation FLiBe Loop under various power levels and operating conditions. The results showed an increasing and decreasing trend in the overall heat transfer with optical thickness with a maximum at optical thickness τ D ≈Ο(1). A forced convection FLiBe loop was constructed for the purpose of studying corrosion and thermal hydraulic phenomena. A vertical cantilever pump is used to achieve up to 5 gpm and 5 m/s flow velocity in hot and cold test sections. A methodology has been established for modeling radiative heat transfer (RHT) in molten salt coolants for the purpose of quantifying the contribution of thermal radiation in conditions relevant to FHR or MSR systems. The comparison between the results obtained for forced and mixed convection showed that in the studied ranges of optical thickness, the maximum effect of RHT on overall heat transfer was 60% improvement observed at z/D = 250 for forced convection; whereas for mixed convection, that value was higher at 170% observed at z/D = 30. The highest effect of RHT on the overall energy distribution was observed at τ D ≈3.4. Negligible effect (<20%) was observed above τ D =18 and below τ D ≈0.002. A CFD code-to-code comparison was conducted using COMSOL Multiphysics at UW-Madison and STAR-CCM at MIT, for a forced and mixed convection laminar flow for a grey medium in a vertical heated using both COMSOL Multiphysics at UW-Madison and STAR-CCM at MIT. The relative error was the largest at τ D =40 for the radiative heat flux, bulk temperature, and Nusselt number. On the other hand, the relative error was the least at τ D =0.002 for the conductive heat flux, bulk temperature, and Nusselt number.

Anderson, Mark↗
DOE OSTI2016

Thermal Desorption of Tritium from Fluoride Salt-Cooled Reactor Materials

The fluoride-salt-cooled high-temperature reactor (FHR) is a high-temperature fission reactor concept involving a liquid fluoride salt coolant, coated particle fuel, and enhanced passive safety mechanisms. Utilizing a low-pressure coolant and high temperature fuel allows for smaller reactor buildings and use of gas Brayton power cycles for higher efficiency in electricity production. One of the fluoride salts being considered for the FHR coolant is 2LiF - BeF{sub 2} (flibe). Tritium is generated frequently in flibe from thermal neutron transmutation of Li-6 and Be-9. Tritium retention in irradiated structural materials presents a challenge to the design of the FHR; high neutron fluence can induce significant material damage and alteration of microstructures to trap tritium in the structure, particularly in carbon-based materials (such as steels). Investigation of tritium transport in flibe salt and structural materials is also critical to evaluating mass balance of tritium in a system for quantifying leakage of tritium to the environment. It is important to understand the uptake of tritium in these proposed core materials since the transport of tritium within an FHR is of critical concern for operational safety due to corrosion, and to prevent tritium release into the environment. The presence of TF corrodes stainless steel by oxidizing chromium. Tritium gas (T{sub 2}) easily diffuses through reactor materials and can be released into the environment through the primary piping and heat exchangers. There is great interest in identifying materials that are compatible with the FHR environment that can be used to getter tritium from the salt, or to inhibit the diffusion of tritium to prevent its release into the environment. As part of the FHR Integrated Research Project (IRP), capsules containing proposed core materials immersed in flibe salt were irradiated at 700 deg. C for 1000 hours (FS-1) and 300 hours (FS-2) in the MIT Research Reactor (MITR). During irradiation, tritium is produced in the salt, and then interacts with the other materials in the capsule. Measurements of the effluent gas during the FHR IRP irradiation experiments indicated that less than 1% of the tritium generated in the salt escaped in the gas phase. It is assumed that most of the tritium was captured, either via diffusion or surface adsorption, on the materials inside the capsule. Tritium frequently adsorbs to sample surfaces in the form of HTO, which requires relatively low temperatures to release (on the order of 100 deg. C). In certain cases, tritium can be bound in the bulk of materials from diffusion or neutron transmutation of lithium and beryllium in the structural materials. Accurate measurement of very small concentrations of hydrogen is difficult; by taking advantage of the radioactive decay of tritium (which emits low-energy beta radiation), several methods become available to detect and quantitatively measure the tritium. The most sensitive method is liquid scintillation counting (LSC), which requires the tritium to be suspended in optically-transparent liquid media that is compatible with scintillating solution, such as water. This measurement technique requires the tritium to be extracted from the material, which can be achieved through leaching into water, thermal desorption, or complete destruction of the sample by acid digestion or pyrolysis. A less destructive method is preferred in order to preserve the specimens, reduce the generation of radioactive mixed wastes, and minimize the carryover of other activation products. For total thermal desorption of tritium from the materials, the tritiated materials must be heated to temperatures in excess of 300 deg. C. Heating the specimens above the irradiation temperature will re-mobilize the tritium and allow it to be captured from the gas stream directly into LSC media. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Hydrogen and deuterium permeation in Hastelloy N

Hastelloy N was chosen as the fluoride salt-contacting structural material for the Molten Salt Reactor Experiment due to its excellent compatibility with the fuel salt FLiBe. FLiBe is currently investigated for several advanced fusion and fission reactor concepts where tritium generation in the FLiBe is anticipated. Knowledge of hydrogen transport properties through Hastelloy N is important to understand how tritium would permeate through this material and result in an unintentional release. In this study, the hydrogen and deuterium permeability, diffusivity, and solubility were measured from 500 to 700 °C at a primary-side pressure of 10 kPa in a well-characterized sample of Hastelloy N. The prepared polycrystalline Hastelloy N had C and O impurities present on the surface. These impurities were investigated using Auger Emission Spectroscopy and Ar depth profiling. Here, the adventitious C was removed upon the first Ar sputter cycle whereas O persisted deeper into the sample. For permeation experiments, applied deuterium pressures ranged from 13 Pa to 100 kPa and deuterium transport remained in the diffusion-limited regime (J ∝ P 0.5 ) throughout the pressure range examined. Two methods are employed to measure the effective hydrogen and deuterium diffusivity: rise and decline. The decline method produced improved statistical model fits for calculating the effective diffusion coefficient compared to the rise method. The resultant transport properties compared well to published values for other nickel alloys.

08 HYDROGEN↗
DOE OSTI2025

Investigation of a U(IV)/U(III) Thermodynamic Reference Electrode for High-Temperature Molten Fluoride Salts

While thermodynamic reference electrodes with known and stable potentials are common in traditional aqueous systems, the high temperature and corrosive environment of a molten fluoride salt makes achieving long term stability with a thermodynamic reference electrode challenging, especially at temperatures of 600°C or higher. In this work, a thermodynamic reference electrode consisting of U(IV)/U(III) in a boron nitride compartment was evaluated for use in FLiBe at temperatures ≥ 600°C. FLiBe used in the study was purified by AlphaTech's proprietary process and characterized by ICP-MS and square wave voltammetry. The free oxide concentration was <2 ppm. Using the purified FLiBe, the U(IV)/U(III) thermodynamic reference electrode was shown to provide a stable, well-defined, and reproducible potential for more than 600+ hours of use in different tests. Moreover, the thermodynamic reference electrode showed a consistent potential with no signs of failure, even after being cooled between tests and then reheated for reuse. Thus, the U(IV)/U(III) reference electrode is suitable for use in rigorous electrochemical studies in molten fluoride salts. As a result, it may be useful as a common standard, facilitating the advancement of nuclear applications such as isotope separation or online monitoring of reactor systems through improved certainty in the measurement of thermodynamic potentials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7 LiF-BeF 2 ) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Electrochemical Behavior of Lanthanum in Molten LiF-NaF-KF 'Flinak' Eutectic Salt

In order to develop the Fluoride High- Temperature Reactor, an emerging Gen-IV reactor concept, a primary coolant cleanup system must be developed to allow online reprocessing of the primary molten fluoride salt. This research is focused on the determination of electrochemical properties of rare earths (REs) in molten fluoride salt related to the FHR. The FHR utilizes two fluoride salts. These fluoride salts are LiF-NaF-KF (46.5-11.5-42 mol%), melting point 454 deg. C, also known as 'Flinak', and LiFBeF{sub 2} (33-67 mol%), melting point 459 deg. C, also known as 'Flibe'. Flibe is selected as the primary coolant salt due to its low neutron cross-section (Li{sup 7}-enriched), superior heat transfer properties, and low melting point. However, due to the beryllium content in Flibe, it is highly toxic to humans and Flinak is preferred for the secondary and tertiary coolant salts. The work presented here is specifically focused on the behavior of lanthanum trifluoride (LaF{sub 3}) in molten Flinak. Furthermore, updated data and analysis regarding the electrochemical behavior of dysprosium trifluoride (DyF{sub 3}) is also presented. Previous work related to this topic was presented at the 2016 ANS winter conference. The goal of this research is to develop a contaminant cleanup system for the primary loop of the FHR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Study of Tritium Transport in Nuclear-Grade Graphite and Molten Fluoride Salt Systems

High-temperature salts are proposed as coolant in recent reactor technologies including Fluoride Salt-Cooled High Temperature Reactor (FHR) with fluoride salt as primary and secondary loop coolant, Molten Salt Reactors (MSRs) with fuel salt, and high-magnetic-field fusion machines with salt coolant blanket. The baseline coolant is {sup 7}Li{sub 2}BeF{sub 4} (flibe) due to its good thermal-hydraulic and neutronic properties. Tritium produced in flibe due to neutron irradiation makes the issue of tritium transport an area of great interest. In the temperature range of 427- 704 deg. C, tritium readily diffuses through metallic reactor structural materials, which makes tritium control and capture a key technical challenge. It is anticipated that tritium can be captured by graphite, as graphite material showed a high affinity for tritium in the Molten Salt Reactor Experiment (MSRE) due to the porous and irregular crystallite microstructure. Two types of nuclear grade graphite which used as core components for FHR: matrix graphite, which encapsulates the TRISO (Tri-structural-isotropic) fuel particles in FHR, and high purity, isotropic nuclear grade graphite IG-110 which can be used as moderator and reflector in reactor core for both fission and fusion reactor. The microstructure of matrix graphite and IG-110 is different from that of matrix graphite, due to the raw material and manufacturing processes. There are three phenomena important for predicting tritium transport in graphite-salt systems: tritium species diffusion in molten flibe; tritium species transport at the salt-graphite interface; and tritium species transport in graphite. In this summary microstructure characterization of IG-110 and matrix graphite, characterization of graphite-salt interface as well as techniques for trace tritium detection in graphite are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2022

Actinides in complex reactive media: A combined ab initio molecular dynamics and machine learning analytics study of transuranic ions in molten salts

The predominant ionic chemistry and the similarity in ionic radius of actinides make it very difficult to structurally distinguish them in liquids. In this work, while ab initio molecular dynamics shows that the f-states clearly affects the electronic properties, their impact on structural properties is not obvious. For the series of trivalent actinides U 3+ , Pu 3+ , Cm 3+ , Cf 3+ , Fm 3+ in molten NaCl and FLiBe, actinide ligand bonds have a higher degree of covalency in NaCl (than in FLiBe), and a higher degree of ionicity in FLiBe. Furthermore, a machine learned classification model can distinguish atomic environments of chemically similar actinides, as long as atoms beyond the first solvation shells are considered. Our work shows that only two types of descriptors are necessary to account for all the fluctuations in heavy metal/molten salt mixtures: The first descriptor represents the electronic state of the heavy metal, while the second encompasses the local coordination environment.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Development and Demonstration of an In-Situ Tritium Scavenger

The objective of this project was to use an inherent feature of salt-cooled reactors towards the capture and removal of tritium from the primary coolant system: the presence of the graphite moderator. The fuel elements fluoride-salt cooled high-temperature (FHR) reactors consist of a graphite matrix that encapsulates TRISO fuel particles, and the total fuel element surface area of Mark I PB-FHR is 7780 m 2 . Therefore the FHR fuel inherently provides removable graphite of high surface area. The goal of this project was to understand the transport of tritium across the salt-graphite interface. The following was achieved: 1. Electrochemical techniques were demonstrated for studying the entry, diffusion, and trapping of hydrogen isotopes in graphite. 2. Electrochemical techniques were used to demonstrate the relationship between galvanic corrosion between graphite and stainless steel 316 and hydrogen uptake in graphite. 3. Electrochemical techniques were demonstrated for studying the dissolution and diffusion of hydrogen isotopes in molten FLiBe. 4. Thermodynamic reference electrodes were tested for FLiBe and FLiNaK. 5. Electrochemical techniques were demonstrated as an in-salt hydrogen thermal desorption measurement, which had previously only been achieved in gas environments. 6. FLiBe exposure of graphite lead to graphite fluorination and an increase of defects, which indicate an increase of reactive carbon sites available for tritium uptake. 7. TRIDENT modeling was used to perform sensitivity analysis of FHR design parameters on tritium uptake fraction in graphite, showing high sensitivity to tritium diffusivity and solubility. 8. A cross-cutting effort with another NEUP-funded project was initiated for application of the SAM mass-transport module to system-scale tritium modeling.

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 OSTI2011

Liquid Wall Chambers

The key feature of liquid wall chambers is the use of a renewable liquid layer to protect chamber structures from target emissions. Two primary options have been proposed and studied: wetted wall chambers and thick liquid wall (TLW) chambers. With wetted wall designs, a thin layer of liquid shields the structural first wall from short ranged target emissions (x-rays, ions and debris) but not neutrons. Various schemes have been proposed to establish and renew the liquid layer between shots including flow-guiding porous fabrics (e.g., Osiris, HIBALL), porous rigid structures (Prometheus) and thin film flows (KOYO). The thin liquid layer can be the tritium breeding material (e.g., flibe, PbLi, or Li) or another liquid metal such as Pb. TLWs use liquid jets injected by stationary or oscillating nozzles to form a neutronically thick layer (typically with an effective thickness of {approx}50 cm) of liquid between the target and first structural wall. In addition to absorbing short ranged emissions, the thick liquid layer degrades the neutron flux and energy reaching the first wall, typically by {approx}10 x x, so that steel walls can survive for the life of the plant ({approx}30-60 yrs). The thick liquid serves as the primary coolant and tritium breeding material (most recent designs use flibe, but the earliest concepts used Li). In essence, the TLW places the fusion blanket inside the first wall instead of behind the first wall.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Tritium Management Loop Design Status

This report summarizes physical, chemical, and engineering analyses that have been performed to support development of a test loop to study tritium migration in FLiBe (2LiF-BeF 2 ) salts. The loop will operate under turbulent flow, and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed, as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated, starting with the development of a Modelica model for a transport through a nickel window into a vacuum, followed by modification of the model for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗