SEARCH · Engineering Papers

Results for “FLUORIDES”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 505 records · Page 28

All sources NASA NTRS DOE OSTI

DOE OSTI2013

Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

Holcomb, David Eugene↗

DOE OSTI2013

Alternatives to the Maroni Process for Tritium Recovery in Fusion Reactors: Avoiding Volatile Hydrogen Fluoride and High-Temperature High-Speed Rotating Machinery

Fusion reactors now on the drawing table, including LLNL’s LIFE reactor, require liquid lithium blankets for cooling and breeding tritium fuel. The tritium fuel must be continuously removed from the blanket so that it can be returned to the reactor for burning, via target manufacturing in the case of LIFE. The baseline process for accomplishing this in the LIFE reactor is the well-known Maroni process which uses high-temperature molten mixed alkali-metal halide salts (LiCl:LiF:LiBr at 530°C) as an extraction solvent. With the generation of hydrogen isotopes, liquid and gaseous acids such as HCl, HF and HBr (and the deuterium and tritium analogs) form in the salt phase, making these solvents very corrosive, with other problems due to volatility. Following extraction of lithium tritide from the molten lithium, the tritide anion must undergo electrochemical oxidation in a high-temperature electrochemical cell to form tritium gas, which can then be separated by stripping in a stream of inert gas such as xenon or helium. The volatility of hydrogen chloride or the tritium analog (HCl or TCl), hydrogen fluoride or the tritium analog (HF or TF) and hydrogen bromide or the tritium analog (HBr or TBr) complicate this separation. An alternative electrolytic process for the conversion of lithium tritide to tritium gas and metallic lithium is based upon the use of a benign lithium hydroxide (LiOH) or lithium carbonate (Li 2 CO 3 ), which is an extremely attractive alternative to the very volatile and corrosive lithium halide electrolyte. The separation of the molten lithium and molten chloride salt phases requires numerous high-temperature (530°C) high-speed centrifugal separators, with bearings and seals exposed to the highly corrosive fluids. Currently, centrifugal contactors are proposed to contact the liquid lithium in which the tritium is bred with a molten salt extraction solvent. These high-speed centrifugal contactors must operate at high temperature for prolonged periods of time without failure. Such rotating machinery, with a highly corrosive molten salt electrolyte, and a liquid lithium phase threatening liquid metal embrittlement is a source of substantial concern. The invention disclosed here is a non-centrifugal contactor capable of operating without any moving parts.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2016

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2 ) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Improving Elemental Ratio Measurements on Uranyl Fluoride with NanoSIMS

Measurements on samples of calcined uranyl fluoride (UO 2 F 2 ) using Raman spectroscopy, electron microscopy with elemental analysis, and secondary ion mass spectrometry (SIMS) indicate four facts. First, UO 2 F 2 undergoes a previously unreported oxidation reaction beginning around 350 °C in air. The resulting oxidation compound is likely similar to U 3 O 8 /UO 3 , but presently represents a novel chemical species. Second, variations in F/U ratio across samples as measured by electron microscopy with x-ray spectroscopy correlate with variations in F/U ratio as measured with the NanoSIMS 50L instrument. Third, simple mixed powders of LaF 3 +U 3 O 8 are not sufficiently homogeneous to act as a high F+U content standard material. Fourth, additional method development for determining the bulk F/U ratio using an alternative technique is necessary to further develop analytical standards.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Thermoelasticity Model for Lithium Fluoride

Lithium fluoride (LiF) is a typical ionic solid, and at the same time a typical polymorphic (multi-phase) material. Its phase diagram contains two solid phases, B1 (rock salt) and B2 (CsCl), which are typical of ionic solids. The phase diagram of LiF was suggested by Smirnov. We have carried out extensive study of the properties of LiF using one of the most recent versions of VASP (Vienna Ab initio Simulation Package) which allows one to carry out calculations on ionic solids; more detail will be presented elsewhere.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Experimental and Modeling Investigation of Overcooling Transients that Include Freezing, in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). Final Report

The goal of this project was an experimental and modeling investigation of overcooling transients that include freezing in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). The most impactful product of this research collaboration between Professor Scarlat’s group at UW Madison and subsequently at UC Berkeley, and Dr. Rui Hu’s group at Argonne National Laboratory was the initial demonstration of SAM application to FHR transients, and the generation of a freezing module for the SAM code. The national laboratory and industry graduate student internships enabled by this project were valuable for workforce development, and also for knowledge transfer between university, national lab, and industry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

SAM Code Enhancement, Validation, and Reference Model Development for Fluoride-salt-cooled High-temperature Reactors

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2022

SCALE Analysis of a Fluoride Salt-Cooled High-Temperature Reactor in Support of Severe Accident Analysis

As part of a US Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with SCALE and MELCOR, SCALE was used for the modeling and simulation of a fluoride salt-cooled high-temperature reactor (FHR). Based on the preconceptual design for a small modular 236 MWth FHR developed by the University of California, Berkeley (PB-FHR-Mk1), a SCALE model of the PB-FHR-Mk1 reactor core was developed. The reactor was modeled at equilibrium state with different fuel compositions in different regions of the reactor. An iterative approach was used to interpolate and mix the burnup-dependent fuel compositions obtained through the depletion calculation of a core slice model. After demonstrating the applicability of SCALE’s multigroup (MG) approach for the simulation of the PB-FHR-Mk1, the resulting equilibrium core was studied in terms of the power profile, the flux profile, temperature reactivity coefficients, and the xenon reactivity. Furthermore, the tritium production rate in the salt coolant was determined, and the dependence of the one-group cross sections on the burnup and location in the reactor core was studied. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the equilibrium core, a zone-wise power profile, temperature feedback coefficients, the tritium production rate, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

SAM Code Development for Transient Safety Analyses of Fluoride-salt-cooled High-temperature Reactors

The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. The molten-salt-cooled pebble-bed reactor, or pebble-bed FHR (PB-FHR) is a promising candidate among advanced nuclear reactor concepts with its improved passive safety characteristics and high thermal efficiency. To support the development and utilization of the SAM code for PB-FHR safety analysis, activities on SAM code enhancements, reference plant model developments, and code validations have been performed in the past a few years to support near-term industry and NRC needs. This report summarizes recent progress under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program in SAM code development and demonstration for transient safety analysis of Fluoride-salt-cooled High-temperature Reactors. SAM capabilities has been significantly enhanced over the years to add FHR specific modeling features, including salt freezing and thawing, spherical core channel and pebble bed core modeling, solid-fluid thermal radiation, tritium transport and general species transport in fluids and solids, and the general code enhancements on solver schemes of point kinetics module and reactivity feedback models. A reference PB-FHR model is developed, based on publicly available information from Kairos Power’s generic FHR design and the University of California, Berkeley (UCB) Mk1 design. A reference reactor model is foundational to the methodologies employed by NRC to verify the adequacy of computer codes and evaluation models. The reference FHR model was utilized for a number of selected FHR design basis accidents, including station blackout, loss of heat sink, loss of flow, transient overpower, and overcooling events.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Safeguards for the Lithium Fluoride Thorium Reactor: A Preliminary Nuclear Material Control and Accounting Assessment

This report was prepared by Oak Ridge National Laboratory (ORNL) for Flibe Energy Inc. (FEI), a US-based advanced reactor company founded in 2011 and headquartered in Huntsville, Alabama. FEI is developing the lithium-fluoride thorium reactor (LFTR), which is a modern two-fluid molten salt reactor (MSR) design operating on a thorium/ 233 U fuel cycle. FEI intends for LFTR to become a self-sustaining clean energy source that can create or breed its own fuel from thorium. Each LFTR is intended to breed enough fissile material to compensate for the amount it consumes. Consequently it would not require fissile replenishment during its operational lifetime. This self-sustaining nature would eliminate the need for uranium enrichment infrastructure to support LFTRs after the first generation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model

In this report we present work performed in Fiscal Year 2022 that demonstrates the modeling and simulation of a fully coupled neutronics thermal hydraulics reference plant model for a fluoride-cooled high-temperature pebble-bed reactor. The multiphysics model is developed on the Nuclear Regulatory Commission’s Comprehensive Reactor Analysis Bundle (BlueCRAB) available on the Idaho National Laboratory’s high-performance computer, which natively and seamlessly couples Griffin, Pronghorn, and the BISON Multiphysics Object-Oriented Simulation Environment based applications. Griffin provides reactor physics capabilities, including depletion to the equilibrium core, k-eigenvalue, adjoint, and transient. The unique direct equilibrium core capability in Griffin is based on a streamline methodology to spatially deplete the pebbles into burnup groups. Pronghorn solves the porous medium equations for the fluid regions and conduction in the solid regions and incorporates a fluidic diode model to simulate the transition from forced to natural convection during accident scenarios. MOOSE modules solves thermal conduction problems for the pebbles and tristructural isotropic in the pebble-bed core, thus providing the fuel and moderator spatial fields for each pebble burnup group. The neutronics feedback relies primarily on fuel, moderator, and reflector temperatures as as well as the FLiBe salt density. Here, we present results for the uncoupled equilibrium core and perform comparisons to equivalent Monte Carlo models. The power distributions and kinetic parameters obtained with Griffin are consistent with those computed with Griffin. We demonstrate a noticeable improvement with the use of discrete ordinates method (SN) transport. The coupled steady-state equilibrium core provides the initial condition for two time-dependent problems: a control rod withdrawal event and an unprotected loss of flow event. In both cases, the reactor design is self-stabilizing and the solutions are consistent with the expected physics. Although this model is prototypical regarding BlueCRAB’s capabilities, its results are consistent with published work by Kairos Power and other research entities. Significant improvements to the model are planned in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Start-up Operation Experience with a Liquid Fluoride Salt Forced Convection Loop

A liquid fluoride salt forced convection test loop was constructed at Oak Ridge National Laboratory. Its unique features include a pebble bed test section, high-temperature instrumentation, a noncontact rotating gas seal for the pump, a silicon carbide flow tube, and a unique inductive heating technique. Initial startup and shake-down testing has been completed. This paper describes how several of the systems performed, highlights those that worked well, and discusses issues that arose during the start-up process. This discussion should be of interest to those who intend to develop molten salt technologies and who are interested in some of the specific techniques used in this experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

SAM Code Development for Source Term Modeling in Fluoride-salt-cooled High-temperature Reactors

The SAM code is under development and supported by DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. These advanced reactor concepts incorporate novel and improved approaches to achieve safety and economic feasibility. Many concepts leverage higher operating temperatures for improved efficiency with a variety of coolants and structural materials to support those needs. Such design choices may also pose unique radiological source term risks. There are continuous efforts to incorporate new physics and phenomena relevant to advanced reactor concepts, including capabilities to enable system-level source term modeling. The Pebble-Bed Fluoride-salt-cooled High-temperature Reactor (PB-FHR) is a promising candidate among advanced nuclear reactor concepts with its improved passive safety characteristics and high thermal efficiency. In addition to past efforts to support the development and utilization of SAM for PB-FHR safety analysis, the species transport modeling capabilities in SAM have been extended to simulate source term phenomena in concepts like the PB-FHR. The PB-FHR concept utilizes pebble-form TRISO fuel and FLiBe salt coolant to provide robust barriers to the release of almost all fission products and radiological source terms. However, tritium poses a unique radiological risk due to its significant production from neutron interactions with 6 Li and 9 Be in the FLiBe as well as its high mobility at elevated temperatures where it can permeate through structural metals. Secondly, the use of graphite at elevated temperatures poses the risk of oxidation damage to any structural components or fuel pebbles that are exposed to accidental air ingress. This report summarizes progress made in modeling these phenomena in SAM, which leverages the system-level multi-physics thermal hydraulic simulation to support an effective engineering-scale source term modeling capability. A tritium transport model is developed to simulate the various transport pathways in FHR and MSR concepts and is presented through verification and validation examples as well as demonstrations simulating experimental test loops and reference plant FHR models. A graphite oxidation model is also introduced to calculate local oxidation rates and is presented with initial validation results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Coupled SAM/Griffin Model of a Reference Fluoride-Salt-Cooled High-Temperature Reactor for Multi-Physics Simulations

A multi-physics coupled simulation model of a reference pebble bed fluoride-salt-cooled high-temperature reactor (PB-FHR) has been developed with SAM and Griffin computer codes for transient safety analysis. The reference problem of a prototypical reactor design serves as the foundation for the U.S. NRC (Nuclear Regulatory Commission) to verify the adequacy of computer codes and evaluation models for specific reactor types. In this work, the previously developed SAM model for PB-FHR primary loop has been updated for the coupled simulation. The updated SAM PB-FHR model includes a 2-D axial symmetric core region and external core components in 0-D/1-D. In addition to the primary loop, a detailed model of the RCCS (reactor cavity cooling system) is added. The 2-D and 1-D domains are tightly coupled using the single-solve approach developed in SAM. In the pebble bed region, the SAM multiscale explicit pebble model is applied to calculate the pebble and TRISO fuel kernel temperatures. The Griffin model used in this work is based on a model developed at Idaho National Laboratory in collaboration with the U.S. NRC. The Griffin neutronics model and SAM thermal hydraulics model is coupled with the Comprehensive Reactor Analysis Bundle (CRAB or alternately BlueCRAB) application. Both steady-state and transient scenarios are simulated to demonstrate the model's suitability for multi-physics simulations of PB-FHR transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Dielectric and Structural Properties of the Hybrid Material Polyvinylidene Fluoride-Bacterial Nanocellulose-Based Composite

In the search for environmentally friendly materials with a wide range of properties, polymer composites have emerged as a promising alternative due to their multifunctional properties. This study focuses on the synthesis of composite materials consisting of four components: bacterial nanocellulose (BNC) modified with magnetic Fe 3 O 4 , and a mixture of BaTiO 3 (BT) and polyvinylidene fluoride (PVDF). The BT powder was mechanically activated prior to mixing with PVDF. The influence of BT mechanical activation and BNC with magnetic particles on the PVDF matrix was investigated. The obtained composite films’ structural characteristics, morphology, and dielectric properties are presented. This research provides insights into the relationship between mechanical activation of the filler and structural and dielectric properties in the PVDF/BT/BNC/Fe 3 O 4 system, creating the way for the development of materials with a wide range of diverse properties that support the concept of green technologies.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Deciphering when Metal Corrosion is Spontaneous in Molten Fluorides Using Potential-Activity Diagrams

The work identifies the conditions for thermodynamically favored spontaneous metal corrosion using potential-activity diagrams tailored for high-temperature molten fluorides. These diagrams provide insights into the thermodynamic phase stability of both solid and dissolved metal species, such as Cr, Cr(II), Cr(III), Ni, Ni(II), Fe, and Fe(II), along with their potential primary oxidizers, including Eu(III), O2, and HF, over a broad range of theoretical F− anion activities. The work further examines the practical implications, prospects, and challenges associated with the construction of these diagrams. The key objective of this project is to pinpoint crucial thermodynamic variables that substantially affect metal corrosion electrochemistry in the context of molten salt nuclear reactor applications.

Materials Science↗

DOE OSTI2017

Molecular determinants of permeation in a fluoride-specific ion channel

Fluoride ion channels of the Fluc family combat toxicity arising from accumulation of environmental F - . Although crystal structures are known, the densely packed pore region has precluded delineation of the ion pathway. Here we chart out the Fluc pore and characterize its chemical requirements for transport. A ladder of H-bond donating residues creates a ‘polar track’ demarking the ion-conduction pathway. Surprisingly, while track polarity is well conserved, polarity is nonetheless functionally dispensable at several positions. A threonine at one end of the pore engages in vital interactions through its β-branched methyl group. Two critical central phenylalanines that directly coordinate F - through a quadrupolar-ion interaction cannot be functionally substituted by aromatic, non-polar, or polar sidechains. The only functional replacement is methionine, which coordinates F - through its partially positive γ-methylene in mimicry of phenylalanine’s quadrupolar interaction. These results demonstrate the unusual chemical requirements for selectively transporting the strongly H-bonding F - anion.

59 BASIC BIOLOGICAL SCIENCES↗