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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Dissolution of Fluoride Salts in Hanford Tank Waste

The Direct Feed High-Level Waste (DFHLW) strategy seeks to bypass the Hanford Waste Treatment and Immobilization Plant Pretreatment Facility while retaining some processing functions to maximize waste feed loading and minimize high-level waste (HLW) waste volume. The DFHLW flowsheet needs leaching, washing, and solids concentration operations either in new or existing tanks. The effectiveness and efficiency of sludge washing has a substantial impact on DST space, mission duration, and the evaporation and low-activity waste (LAW) treatment operations required by these large wash-water additions. One target species requiring washing is fluoride. The HLW glass composition limits for fluorine drive operations to dissolve fluoride-bearing salts into the LAW fraction and thereby maximize waste loading in HLW glass. The fluoride in many high-level wastes at Hanford is predominantly in the form of fluoride-salt precipitates: villiaumite (NaF), kogarkoite (Na 3 FSO 4 ), and natrophosphate (Na 7 F(PO 4 ) 2 ·19H 2 O). Fluoride produces melter off-gas that creates corrosion risk in the off-gas system piping, while the sulfate and phosphate in the fluoride double salts kogarkoite and natrophosphate can be detrimental to glass waste loading. Fluoride salts are sparingly soluble, with solubilities ranging from approximately 40 to 130 kg per kL of pure water, and the dissolution kinetics of the three fluoride salts are not well known. Unexpected delays in a tank dissolution process could be encountered as a result of the lack of information about dissolution rates. In addition, if the double salts show transient non-stoichiometric dissolution of fluoride versus phosphate or sulfate, unexpectedly high concentrations of one of these other constituents could be produced. Washington River Protection Solutions authorized Pacific Northwest National Laboratory to collect the available data for fluoride salt dissolution rate, provide a scoping estimate of dissolution time if possible, and identify gaps in the understanding and predictive capability for estimating dissolution time. Open literature and Hanford reports were reviewed to document, understand, and (where possible) evaluate limitations on fluoride salt equilibria and dissolution kinetics, including both mass transport and surface reaction rate. Scoping estimates of dissolution time were made for mass-transfer-controlled dissolution of spherical particles of fluoride salts suspended in liquid. This is not the only potential governing mechanism; dissolution could be substantially slower if the surface reaction rate (the rate of release of ions from the surface) is the controlling mechanism. When the minimum amount of water for complete dissolution is used and the slip velocity between the liquid and suspended particles is less than or equal to the terminal settling velocity, the estimated mass-transfer rates allow 0.1 mm particles of fluoride salt to dissolve in minutes at 25 °C in water containing no other dissolved salts. Much larger solids, such as the 6-mm chunks that have been seen in heels, could take a few hours to more than a week to dissolve. The actual dissolution times will depend strongly on the actual slip velocity, the extent of particle suspension, constraint by surface reaction rates, the ratio of solvent to solid, and the presence of common ions that shift the solubility equilibria to restrict dissolution of fluoride salts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Thermochemical Property Measurements of FLiNaK and FLiBe in FY 2020

The technical bases and methods developed at Argonne to measure thermochemical properties of molten salt mixtures are summarized and results provided to show the precision achieved. These methods include measurements of phase transition temperatures and heat capacity at temperatures to approximately 700 °C for eutectic mixture of 46.5-11.5-42.0 mol% LiF-NaF-KF (FLiNAK) and 67-33 mol% LiF-BeF 2 (FLiBe). Details of procedures, calibrations and operation of instruments are discussed and results are compared with values available in the literature. Additional discussions address the sources of error and estimated uncertainties in the reported values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2021

Precision of Property Measurements with Reference Molten Salts

Thermochemical and thermophysical properties of reference molten salt systems are being measured to establish the precisions achievable with current state-of-the-art methods including repeatability within a laboratory and reproducibility between laboratories using the same and different methods. Several laboratories across the United States are analyzing salt samples from single source batches of two salts to provide independent measurements of property values and the precisions of different measurement techniques in an interlaboratory study referred to as the Interlaboratory Salt Study (ISS). The compositions of the two salts produced for the study were reported to be eutectic LiF-NaF-KF (46.5-11.5-42 mol %) and NaCl-KCl (50-50 mol %). These salts are referred to herein as ISS FLiNaK and ISS NaCl-KCl. Thermochemical properties being measured at Argonne include thermal transitions and heat capacity by differential scanning calorimetry. Thermophysical properties being measured at Argonne include liquid density and surface tension by using a buoyancy densitometer and thermal diffusivity by using a laser flash analyzer. Thermal conductivity is calculated from measured thermal diffusivity, density and heat capacity values. The results of property measurements made at Argonne using these inter-laboratory salt study salts are presented here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Analytical Results for Prepared UIUC F 7 LiNaK

Source materials including 275 g of 7 LiF-NaF-KF (46.5-11.5-42.0 mol %) and 10 g each of several reagent salts were prepared at Argonne for shipment to the University of Illinois at Urbana Champaign. The F 7 LiNaK was packaged in a single jar and reagent salts were placed in small vials. The mass of each shipped salt is listed in Table 1. This report contains details of the preparation procedure and results of elemental composition measurements using inductively coupled plasma-mass spectrometry (ICP-MS) and inductively coupled plasma-optical emission spectrometry (ICP-OES) and phase analyses using X-ray diffraction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Electroanalytical Sensors for Liquid Fueled Fluoride Molten Salt Reactor

ThorCon is developing a thermal thorium / uranium molten salt fueled 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. On-site fuel processing is limited to adding make-up fuel, and adding beryllium to maintain redox balance in the fuel salt. The fluoride salt is an excellent flux that will strip the protective oxygen away from all metal surfaces it contacts. In order to protect the metal surfaces in the primary loop, no free fluorine is allowed. Maintaining a ratio of UF 4 to UF 3 provides a chemical buffer that can absorb any free fluorine released during the fission process before it can harm the structure. Periodic additions of metallic beryllium to the fuel salt will react with UF 4 to form BeF 2 and UF 3 , which will maintain the desired UF 4 /UF 3 ratio. A means to measure the UF 4 /UF 3 ratio is important to maintain the fuel salt in a non-corrosive state and could also provide additional insight into the corrosion behavior and chemical constitution of the system. A promising measurement technique is based on electrochemistry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Thermal Property Characterization of Molten Salt Reactor–Relevant Salts

The thermochemical and thermophysical properties of molten salt systems are necessary input for the design, licensing, and deployment of molten salt reactor concepts. The thermochemical and thermophysical property measurement capabilities being used and developed at Oak Ridge National Laboratory (ORNL) include liquid density, volumetric thermal expansion, viscosity, vapor pressure, phase behavior, and heat capacity and thermal conductivity measurement capabilities. Properties of eutectic LiF-NaF-KF (FLiNaK) and NaCl-KCl are being measured to validate the measurement techniques to be used to provide data to reactor developers and to better quantify the uncertainty of measurements. This report summarizes the status of thermal conductivity, viscosity, specific heat capacity, and density. A method for property estimation using Redlich Kister expansion is also described. This technique will be used to help validate measured properties and to guide future experiments by targeting mixtures or components for which data are uncertain or lacking.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Thermochemical Property Measurements of FLiNaK and FLiBe in FY 2020

The technical bases and methods developed at Argonne to measure thermochemical properties of molten salt mixtures are summarized and results provided to show the precision achieved. These methods include measurements of phase transition temperatures and heat capacity at temperatures to approximately 700°C for eutectic mixture of 46.5-11.5-42.0 mol% LiF-NaF-KF (FLiNaK) and 67-33 mol% LiF-BeF 2 (FLiBe). Details of procedures, calibrations and operation of instruments are discussed, and results are compared with values available in the literature. Additional discussions address the sources of error and estimated uncertainties in the reported values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Understanding FLiNaK Salt Intrusion Behavior on Nuclear-Grade Graphite via Neutron Tomography

Graphite is an essential material as a neutron moderator in molten salt reactors (MSRs). To understand the impact of salt on the graphite structure to develop structural materials for MSRs, molten salt intrusion behavior on nuclear-grade graphite was studied. The graphite samples, IG-110 and PCEA, were tested for infiltration with LiF-NaF-KF (FLiNaK) at 750°C, 5 bar for 12 h, and, after the intrusion experiment, the graphite was analyzed by neutron imaging. The graphite and Li from FLiNaK showed great contrast in the neutron attenuation coefficient. Thus, the salt behavior in the graphite structure has been visualized for the first time without damaging the sample. The 3D image of the graphite was reconstructed after a neutron computed tomography scan, and the average salt coverage distribution of the XY surface in different depths was obtained from the reconstructed 3D images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Molten Salt Loop Testing of Sensors and Off-Gas Components: FY23 progress

The Liquid Salt Test Loop (LSTL) at the US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) was developed to demonstrate technology for high-temperature fluoride salt systems (Figure 1). The LSTL is primarily constructed using Inconel 600 alloy and operates at temperatures of up to 700°C. The facility is loaded with 165 kg of LiF-NaF-KF salt (FLiNaK). This salt provides a relevant test environment for de-risking technology while avoiding the costs and hazards associated with beryllium-based or uranium-bearing salts. FLiNaK is also an advantageous salt for the secondary side of molten salt reactors. The facility’s major components include a centrifugal pump for salt circulation, an air-based heat exchanger to reject heat, a suite of instrumentation, and trace heating to prevent salt freezing. Additional heating is available through an induction heater rated at 200 kW. The relatively large heating and cooling capability enables the formation of a temperature gradient across the loop (i.e., a hot and a cold side), which is important for chemistry and corrosion studies. The LSTL is a unique US capability for high-temperature molten halide salt testing. Although some efforts are underway at universities, the LSTL’s scale, co-located purification system, and relatively large power differentiates it from other testing systems. Furthermore, unlike efforts within industry, access to the DOE-supported facility and communication of results, which are generally disseminated publicly, result in a broad significance in the molten salt reactor community.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Molten Salt Loop Operational Experience and Test Campaigns in FY24

The Facility to Alleviate Salt Technology Risks (FASTR) at the US Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) was developed to demonstrate technology for high-temperature chloride salt systems (Figure 1). FASTR is primarily constructed using alloy C-276 and is designed to operate at temperatures of up to 725°C. The facility is loaded with 250 kg of NaCl-KCl-MgCl 2 salt. This salt provides a relevant test environment for de-risking technology while avoiding the costs and hazards associated with beryllium-based or uranium-bearing salts. The facility’s major components include a centrifugal pump for salt circulation, an air-based heat exchanger to reject heat, a suite of instrumentation, and trace heating to prevent salt freezing. The salt was purified in 2020 and 2022, and the pumped loop first operated in December 2022. FASTR is a unique US capability for high-temperature molten halide salt testing. FASTR’s scale, co located purification system, and relatively large power (465 kW) differentiates it from other testing systems. Furthermore, access to the DOE-supported facility and efficient communication of results— which are generally disseminated publicly—distinguish FASTR as being broadly significant throughout the molten salt reactor community. FASTR is similar to ORNL’s Liquid Salt Test Loop (LSTL), although FASTR contains chloride-based salt instead of the fluoride-based salt (LiF-NaF-KF) found in LSTL. Furthermore, FASTR is approximately 2× larger than LSTL in terms of pipe size and length, power, salt volume, flow rate, and number of thermocouples. The LSTL first operated in 2016. At the end of FY23, there was a suspected gas leak in the LSTL that halted operation. At the start of FY24, a leak in the LSTL pump’s tank gas space was confirmed. Because the gas-space leak prevented operation of LSTL, FY24 efforts were focused on operation of FASTR. This report summarizes the progress made during FY24 in support of the DOE Office of Nuclear Energy (DOE-NE) work package, AT-24OR070202 Salt Loop and Capability for Testing Sensors and Off Gas Components.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Friction and wear of nuclear graphite exposed to molten FLiNaK salt environment

This report formally documents the completion of the Advanced Reactor Technologies Level 3 Milestone (M3TG-25OR0501104), “Initiate graphite wear studies of graphite samples exposed to molten salt environment,” due August 1, 2025. The report summarizes the ongoing activities aimed at characterizing the friction and wear behavior of graphite in molten LiF–NaF–KF (FLiNaK) salt environment. The wear and friction behavior of self-mated ET-10 nuclear graphite were tested at varying temperatures (550°C and 650°C), contact loads (40 and 80 N) and sliding speeds (1 and 10 mm/s) in a controlled argon environment. The results were compared to the wear and friction behavior of the same graphite material in a dry argon environment. This report presents initial studies on the friction and wear behavior of graphite in sliding contact mode, with future work focused on investigating the graphite matrix materials, as well as rolling and impact contact modes. The outcomes of this report could provide insights into assessing the integrity of the graphitic components of pebble-bed fluoride salt–cooled high-temperature reactors (PB-FHRs) to enhance their safe operation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1975

Research for preparation of cation-conducting solids by high-pressure synthesis and other methods

It was shown that two body-centered-cubic skeleton structures, the Im3 KSbO3 phase and the defect-pyrochlore phase A(+)B2X6, do exhibit fast Na(+)-ion transport. The placement of anions at the tunnel intersection sites does not impede Na(+)-ion transport in (NaSb)3)(1/6 NaF), and may not in (Na(1+2x)Ta2 5F)(Ox). The activation energies are higher than those found in beta-alumina. There are two possible explanations for the higher activation energy: breathing of the bottleneck (site face or edge) through which the A(+) ions must pass on jumping from one site to another may be easier in a layer structure and/or A(+)-O bonding may be stronger in the cubic structures because the O(2-) ion bonds with two (instead of three) cations of the skeleton. If the former explanation is dominant, a lower activation energy may be achieved by optimizing the lattice parameter. If the latter is dominant, a new structural principle may have to be explored.

Goodenough, J. B.↗
NASA NTRS1975

Alkali oxide-tantalum, niobium and antimony oxide ionic conductors

The phase equilibrium relations of four systems were investigated in detail. These consisted of sodium and potassium antimonates with antimony oxide and tantalum and niobium oxide with rubidium oxide as far as the ratio 4Rb2O:llB2O5 (B=Nb, Ta). The ternary system NaSbO3-Sb2O4-NaF was investigated extensively to determine the actual composition of the body centered cubic sodium antimonate. Various other binary and ternary oxide systems involving alkali oxides were examined in lesser detail. The phases synthesized were screened by ion exchange methods to determine mobility of the mobility of the alkali ion within the niobium, tantalum or antimony oxide (fluoride) structural framework. Five structure types warranted further investigation; these structure types are (1) hexagonal tungsten bronze (HTB), (2) pyrochlore, (3) the hybrid HTB-pyrochlore hexagonal ordered phases, (4) body centered cubic antimonates and (5) 2K2O:3Nb2O5. Although all of these phases exhibit good ion exchange properties only the pyrochlore was prepared with Na(+) ions as an equilibrium phase and as a low porosity ceramic. Sb(+3) in the channel interferes with ionic conductivity in this case, although relatively good ionic conductivity was found for the metastable Na(+) ion exchanged analogs of RbTa2O5F and KTaWO6 pyrochlore phases.

Roth, R. S.↗
NASA NTRS1979

Electrochemical fluorination of trichloroethylene and N, N-dimethyltrifluoroacetamide

Fluorination of trichloroethylene and N, N-dimethyltrifluoroacetamide was carried out on a laboratory scale in an advanced Simons type electrochemical apparatus which could be operated automatically from ambient to 50 psi pressure. A variety of fluorine-substituted products are formed, depending upon electrolysis conditions and concentrations of reactant relative to the NaF, KF, HF electrolyte. A new reaction mechanism of electrochemical fluorination of trichloroethylene is proposed. The solvency-to-fluorine content relationship of fluorinated N, N-dimethyltrifluoroacetamide is described.

Hsu, L. C.↗
NASA NTRS1979

Electrochemical fluorination of trichloroethylene and N, N-dimethyltrifluoroacetamide

The paper presents the results of experiments concerning the fluorination of trichloroethylene and N, N-dimethyltrifluoroacetamide carried out on a laboratory scale in an advanced 'Simons' type electrochemical apparatus which could be operated automatically from ambient to 50 psi pressure. It is shown that a variety of fluorine-substituted products are formed, depending upon electrolysis conditions and concentrations of reactant relative to the NaF, KF, HF electrolyte. A new reaction mechanism of electrochemical fluorination of trichloroethylene is proposed. Finally, the solvency-to-fluorine content relationship of fluorinated N, N-dimethyltrifluoroacetamide is described.

Hsu, L.-C.↗