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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

An electrochemical study of hydrogen in molten 2LiF-BeF 2 (FLiBe) with addition of LiH

The chemical and transport behavior of hydrogen isotopes in molten 2LiF-BeF 2 (FLiBe) is of interest for the design of tritium management systems in nuclear fission and fusion reactors that use FLiBe. The chemical reaction of LiH with FLiBe is used to introduce hydrogen in the molten salt and electrochemical methods are used for in situ studies of hydrogen in FLiBe. LiH reacts with molten FLiBe to generate an electroactive species whose voltammetry peak is proportional to the added quantity of LiH. The cyclic voltammetry reaction potential of 2.009 ± 0.050 V vs Be/Be 2+ and the electron exchange of n = 0.8 ± 0.5 are consistent with the one-electron oxidation of dissolved H in the zero valence state, H o to H + . The concentration of H o is estimated by linear sweep voltammetry at 60–80% of the hydrogen introduced by the reaction of 0.42 mol% LiH with FLiBe and after eleven hours it remains above 40%. It is postulated that covalent BeH 2 is formed in FLiBe upon LiH addition, as a FLiBe-soluble quasi-stable intermediate product. Furthermore, the results provide an evaluation of LiH as a means of introducing dissolved hydrogen in FLiBe, enabling electrochemical methods as tools to advance the understanding of the chemistry of hydrogen isotopes in FLiBe.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The LIBRA Experiment: Investigating Robust Tritium Accountancy in Molten FLiBe Exposed to a D-T Fusion Neutron Spectrum

The Liquid Immersion Blanket: Robust Accountancy (LIBRA) experiment will be a first-of-a-kind experiment to explore and develop the liquid immersion blanket (LIB) concept. The LIB is a radically simple molten–LiF-BeF2 (FLiBe)–salt tritium breeding blanket for deuterium-tritium (D-T)–fueled fusion power plants (FPPs) achieving a high tritium breeding ratio (TBR) in neutronics models. However, tritium breeding in FLiBe is inherently difficult to study experimentally. As a result, the coupled issues of FLiBe radiochemistry and tritium (T) transport are poorly understood. LIBRA approaches this challenge by simulating an FPP blanket environment using a D-T neutron generator and ≈ 1000 kg of FLiBe. LIBRA will investigate T breeding, containment, and extraction, coupled with FLiBe redox control and radiochemistry. The primary goal of LIBRA is to demonstrate robust T accountancy in blanket prototypical conditions. Here, T accountancy encompasses accurate predictions of T breeding in the FLiBe; detection and measurement of all T bred in LIBRA; and speciation of the T extracted from the FLiBe. Initial neutronics simulations of LIBRA indicate that a global TBR of ~ 1 is possible, where the TBR is defined as the number of tritons bred and extracted from FLiBe relative to the number of neutrons produced by D-T fusion reactions in the neutron generator. In this paper, we present the LIBRA concept and its scientific goals in the context of T breeding experiments. We also consider the potential impact of the LIB on the future fusion power industry, motivating further development of FLiBe-based T breeding research activities such as LIBRA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Sensitivity/uncertainty analyses comparing LR-0 reactor experiments containing FLiBe salt with models for molten-salt-cooled and molten-salt-fueled reactors

Here, critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Rež (RC Rež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7 Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7 Li. Other smaller but potentially significant contributions come from 19 F, 6 Li, and other 7 Li reactions. Similarity comparisons of the RC Rež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233 U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19 F and covariance between reactions such as 233 U fission and 233 U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Compatibility of Type 304H Stainless Steel in Static Molten FLiBe for Inertial Confinement Fusion Reactors: Role of Impurities and Redox Control

Molten fluoride salts, such as FLiBe (LiF-BeF2), are promising candidates for tritium breeding and heat transfer in fusion reactors, but corrosion of structural materials remains a major challenge. This study investigates the corrosion behavior of austenitic stainless steel 304H in purified and NiF2-containing FLiBe at 500°C and 600°C, focusing on the effects of impurities and redox control. Exposure to purified FLiBe resulted in the concurrent depletion of Cr, Mn, and Fe, with corrosion at 500 °C dominated by the combined oxide formation and elemental dissolution, while at 600°C elemental depletion was predominant. The addition of a controlled NiF2 impurity significantly accelerated corrosion at both temperatures, demonstrating the sensitivity of 304H to the salt redox state. Beryllium additions were effective in mitigating corrosion for both baseline and NiF2-containing FLiBe; minimal depletion of Cr, Mn, and Fe occurred with Be additions as low as 2.5 mg (147 wppm), and no NiBe intermetallics formed at 5 mg (294 wppm), indicating that small Be inventories can provide substantial protection without deleterious phase formation. Thermodynamic equilibrium and coupled thermodynamic-kinetic analyses at the salt-alloy interface suggested low corrosion rates in systems with limited hydrogen fluoride (HF) generation, highlighting the importance of salt redox control. Estimates for a Be addition rate were calculated for the HYLIFE-II fusion reactor that can mitigate corrosion-induced degradation, assuming complete conversion of tritium to tritium fluoride (TF). Overall, 304H shows reasonable compatibility with FLiBe under optimized redox conditions. These results provide quantitative guidance for material selection and salt management in fusion blanket and heat exchanger systems and motivate validation under flowing, nonisothermal, and irradiated conditions.

Pillai, Rishi [ORNL] (ORCID:0000000243688197)↗
DOE OSTI2025

Assessing Stainless Steel Compatibility in Flowing FLiBe Salt

To address the considerable interest in LiF-BeF 2 (FLiBe) compatibility for fission and fusion reactor applications, static and flowing compatibility experiments were conducted to assess the compatibility with type 316H stainless steel. In static testing at 550° and 650 °C, small mass changes were measured and posttest characterization of the FLiBe showed increased levels of Fe, Cr, Ni and Mn in the salt. Adding Be in the static salt test reduced the dissolution of Fe and Ni. Further, an initial assessment of mass transfer in flowing FLiBe without a Be addition was conducted using a monometallic 316H thermal convection loop (TCL) operated for 1000 h with a peak temperature of 650 °C. Similar to prior results in flowing FLiNaK salt, the 316H specimens exhibited small mass losses in the hot leg. Posttest characterization of the 316H specimens suggested Cr surface depletion in the hot and cold legs and possibly Fe deposition in the cold leg. To further understand this behavior, Cr and Fe dissolution was measured in static FLiBe at 550–650 °C.

316H stainless steel↗
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↗
DOE OSTI2022

Corrosion of 316H Stainless Steel Specimens in Two FLiBe (LiF-BeF 2 ) Salt Batches

This milestone was originally envisioned for completion in FY21 and was delayed due to the COVID-19 response and difficulties in fabricating fluoride salts. To complete the milestone, commercial FLiBe was compared to a batch of FLiBe produced by the conventional hydrofluorination process. However, the batch was stopped early due to an HF leak and the salt did not undergo the final H2 sparging. Subsequent static compatibility testing of 316H specimens in 316H capsules resulted in small mass losses for the commercial FLiBe and much larger mass losses for the 2nd batch of FLiBe.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Constant-potential molecular dynamics simulations of molten salt double layers for FLiBe and FLiNaK

We report the results of constant-potential molecular dynamics simulations of the double layer interface between molten 2LiF–BeF 2 (FLiBe) and 23LiF–6NaF–21KF (FLiNaK) fluoride mixtures and idealized solid electrodes. Employing methods similar to those used in studies of chloride double layers, we compute the structure and differential capacitance of molten fluoride electric double layers as a function of applied voltage. The role of molten salt structure is probed through comparisons between FLiBe and FLiNaK, which serve as models for strong and weak associate-forming salts, respectively. In FLiBe, screening involves changes in Be–F–Be angles and alignment of the oligomers parallel to the electrode, while in FLiNaK, the electric field is screened mainly by rearrangement of individual ions, predominantly the polarizable potassium cation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Materials Assessment for FLiBe Fusion Blankets

This evaluation was prepared at the request of DOE-FES and expands on an earlier evaluation prepared by Wilson, Wiffen and Keiser in 2015 (attached as Appendices A and B). Additional useful information for materials in fission molten salt reactor systems that use FLiBe can be found in a prior NRC report [Busby 2019]. The purpose of this report is to review the considerations that will determine the fusion blanket structural materials that will be suitable for use in a FLiBe breeder-coolant fusion power plant (FPP). Identifying a fusion-relevant structural material that is sufficiently compatible with FLiBe is a key issue to be addressed. The constraints and trade-offs made in structural materials selection include the usual power generation requirements such as mechanical properties, fabricability and durability, which assumes reasonable compatibility with the working fluid. For an FPP, additional constraints include neutron economy (breeding ratio), resistance to irradiation damage and low activation (i.e. waste disposal) characteristics.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Application of voltammetry for quantitative analysis of chromium in molten 2LiF-BeF 2 (FLiBe) salt

The electrochemical behavior of Cr 2+ in molten 2LiF-BeF 2 (FLiBe) salt at temperatures between 500 °C and 700 °C was studied using cyclic voltammetry (CV) and linear sweep voltammetry (LSV). LSV has been used to establish a stable reference electrode with the Be/Be 2+ couple. The ability to distinguish between common corrosion products, such as Cr2+ and Fe 2+ , with CV has been demonstrated using the Be/Be 2+ reference reaction. Methods for high-precision, in-situ quantitative measurement of Cr 2+ concentration are developed. Models for predicting the concentration of Cr 2+ and its diffusion coefficient are approximated by relating faradaic cathodic peak current density to Cr 2+ concentration at 600 °C and 650 °C. Here this study provides the basis for the development of a sensor for on-line quantification of corrosion products that improves understanding of materials performance in the molten FLiBe salt.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Thermophysical properties of FLiBe using moment tensor potentials

Fluoride salts are prospective materials for applications in some next-generation nuclear reactors and their thermophysical properties at various conditions are of interest. Experimental measurement of the properties of these salts is often difficult and, in some cases, unfeasible due to challenges from high temperatures, impurity control, and corrosivity. Therefore, accurate theoretical methods are needed for fluoride salt property prediction. In this work, we used moment tensor potentials (MTP) to approximate the potential energy surface of eutectic FLiBe (66.6% LiF – 33.3% BeF2) predicted by the ab initio (DFT-D3) method. Here, we then used the developed potential and molecular dynamics to obtain several thermophysical properties of FLiBe, including radial distribution functions, density, self-diffusion coefficients, thermal expansion, specific heat capacity, bulk modulus, viscosity, and thermal conductivity. Our results show that the MTP potential approximates the potential energy surface accurately and the overall approach yields very good agreement with experimental values. The converged fitting can be obtained with less than 600 configurations generated from DFT calculations, which data can be generated in just 1200 core hours on today's typical processors. The MTP potential is faster than many machine learning potentials and about one order of magnitude slower than widely used empirical molten salt potentials such as Tosi/Fumi.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The corrosion effects of neutron activation of 2LiF-BeF 2 (FLiBe)

The redox effect of activation reactions in molten salts has been previously qualitatively defined but never quantified. Here, we quantify the corrosion effects of neutron activation of FLiBe. Rates of change of activation products are computed for FHR and ARC. The impact of activation reactions on salt chemistry is quantitatively compared to other environmental effects of similar consequences on the salt. The results are applied to a redox control case-study and an irradiation-corrosion experiment. The results are presented in a manner useful to the reactor chemist and designer concerned with corrosion and chemistry control and tritium management in irradiated FLiBe, and design of salt-irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Evaluation of Corrosion Behavior of Various Fe- and Ni-Based Alloys in Molten Li 2 BeF 4 (FLiBe)

Reliable performance of structural alloys is essential for the successful implementation of Generation-IV fluoride salt–cooled high-temperature reactors (FHRs). Most FHR designs are considering molten salt (2LiF-BeF 2 ), or FLiBe, as a primary coolant or fuel carrier. The main corrosion mechanism for alloys exposed to molten fluoride salts is the selective dealloying of active alloying elements. Alloy composition has a significant effect on their high-temperature mechanical properties, but also affects their corrosion behavior. Although Hastelloy-N and its variants show good corrosion resistance compared to higher Cr-containing Ni- or Fe-based alloys, the mechanical properties of these alloys degrade quickly at temperatures above ~600°C. Twelve Ni-based or Fe-based alloys were selected due to their high temperature stability or their low Cr alloy composition and tested for their corrosion behavior in FLiBe. The results show that the mode and the extent of alloy degradation by selective dissolution mechanism corelates well with the overall alloy composition, and not just the concentration of active elements. It was found that there was good correlation between weight loss of the tested alloys and the ratio of major active elements (Cr, Mn) to that of the more noble alloying elements (Ni, Mo).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Operation Data from the UW Natural Circulation FLiBe Flow Loop

Operating measurements from three multi-week test campaigns of the natural circulation FLiBe loop. This system is investigating thermal hydraulic behavior of the molten salt BeF_2 -LiF (33 - 67 mol %). The system behaves in a stable fashion, but shows unusual local transients, such as flow break-down in the riser and thermal jumps at the cooler exit. Heat transfer shows promising trends that FLiBe will behave as a normal heat transfer fluid, as long as salt purity can be maintained. The most recent test shows heat transfer degradation that is likely a result of oxides and impurities. The data archive contains as-built dimensions, Matlab analysis codes, and the raw data files.

42 ENGINEERING↗
DOE OSTI2021

FLiBe Capsule Testing of 316H

This milestone was originally envisioned for completion in FY20 and was delayed due to the COVID-19 response and lack of FLiBe salt. In March 2021, FLiBe salt was received from Kairos Power. The nine capsule experiments with type 316H stainless steel specimens and 316 capsules were completed at 550° and 650°C for 1000 h to bracket the conditions expected in the subsequent thermal convection loop (TCL) experiment. In three capsules, Be metal was added for redox control. The specimen mass losses were low, similar to those observed previously for FLiNaK salt and the addition of Be reduced the average mass loss. Initial characterization suggested that some reaction with Be may have occurred. Therefore, Be was not included in the TCL experiment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Tritium Extraction from FLiBe using a Permeation Against Vacuum (PAV): Impact of Surface Potentials, Currents, and Coupling PAV to Sacrificial Anodes

Savannah River National Laboratory (SRNL) is exploring a concept of pairing a LiF-BeF2 (FLiBe) immersed Permeation Against Vacuum (PAV) tritium extraction system with a sacrificial anode. Several fusion concepts utilize FLiBe for tritium breeding, and breeding of tritium would occur from neutron interactions with the Li component, creating corrosive TF and/or T2. In absence of a redox buffered system, the TF would be able to attack containment materials until neutralized. One way to neutralize TF is to contact it with an active metal such as Be or Li which would preferentially combine with the fluorine in the TF to form BeF2 or LiF, respectively, and resulting in T2. When the active metal (e.g., Be) is oxidized, it will provide electrons, when T in TF is reduced and forms T2, it needs electrons. SRNL proposes to electrically link these two reactions and with the PAV to help drive the reactions faster than on their own and potentially increase T2 dissociation. In a soluble redox salt buffered system, it is expected that with sufficient buffering, T2, may form more rapidly and the core components would be more protected, but the soluble redox salt buffer would need to be occasionally regenerated, which could be accomplished by again contacting the salt with an active metal such as Be or Li, although in this case the cathodic reaction would be electrons coming from the shift in oxidative state of the redox buffer. The exploration of potential and current effects on the PAV and impacts of redox agents will be discussed.

Olson, Luke C.↗
DOE OSTI2018

FLiBe Radiative Heat Transfer

The liquid fluoride salt, FLiBe (2LiF-BeF2), has been proposed as a working fluid for Fluoride-Salt-Cooled High-Temperature Reactor (FHR) due to its desirable heat transport properties at high temperatures (500-700°C) as well as compatibility with graphite and several metals [1]. Due to the 4th power temperature dependence and the high temperature differences present in the various FHR heat exchangers (TCHX, DHX, and CTAH), radiative heat transfer in molten salt FLiBe may be important compared to natural or forced convection. Furthermore, because of the strong coupling between the temperature and velocity profiles, emission and absorption of radiation within the salt may significantly impact the thermal and hydrodynamic behavior of the fluid in the natural circulation systems of FHR. The present work investigates the significance of radiative transfer within the salt on the overall thermal energy conservation equation, and consequently the overall convective and radiative heat transfer coefficient. The purpose is to understand the radiative heat transfer effect on the steady-state energy equation for a moving salt in a 2D channel by computing a total Nusselt number that incorporates both radiation and convection. Moreover, the evolution of the thermal entrance region is of a particular importance because of the high heat transfer coefficients attainable and the effect on the potential onset of various salt physical phenomena, like freezing and supercooling, as well as on the design of the passive decay heat removal systems of FHR, which are mostly buoyancy-driven pipe flow systems.

Derdeyn, Will↗
DOE OSTI2019

Natural circulation FLiBe loop overview

A new large natural circulation FLiBe loop has operated for the first time in 40 years. Reaching temperatures above 700 C with stable flow rates around 5 cm/s, the loop exhibits a range of flow phenomena pertinent to high temperature natural circulation safety systems for nuclear reactors. This paper discusses the system and several operating conditions spanning a wide range of behavior to aid the development of other systems, simulation benchmarks, or system-codes. Analysis of heat transport phenomena indicates that standard Nusselt number correlations apply to a first approximation. More accurate salt-freezing models may be necessary for air-cooled salt-systems due to the insulating affect of the low thermal-conductivity frozen salt layer on the cold pipe walls.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗