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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

The Electrochemical Behavior of Tellurium Tetrachloride in LiCl-KCl Eutectic Molten Salt at 450 ºC

Cyclic voltammetry CV), square wave voltammetry (SWV), and electrochemical impedance spectroscopy measurements were carried out in 1 and 5 wt% TeCl 4 containing LiCl-KCl eutectic molten salt at 450 °C using tungsten and glassy carbon as working electrodes. Reduction of Te(IV) to Te(0) occurred in three steps at potentials more negative than -0.4 V Ag/AgCl . The Te(0) was further reduced to Te 2- at more negative potentials. Formation of Te 2- was observed at potentials more negative than -2.2 V Ag/AgCl . Diffusivities of Te 4+ , Te 2+ , and Te 2- and their formal potentials were estimated from the CV data. The diffusion coefficients of Te 4+ , Te 2+ and Te 2- were 0.9 × 10 -5 , 3.8 × 10 -5 , and 1.5 × 10 -5 cm 2 s -1 , respectively in the 1 wt% TeCl 4 containing LiCl-KCl molten salt. The diffusivity of Te species generally decreased with increase in the concentration of TeCl 4 .

Electrochemistry↗
DOE OSTI2021

Communication—Mg 2+/0 as a Reliable Reference Electrode for Molten Chloride Salts

In this report, we describe the use of the Mg 2+/0 couple as a reference in molten MgCl 2 -NaCl. We demonstrate that this redox couple provides a robust and reliable reference potential over a range of melt temperatures (500 °C–675 °C). Finally, we demonstrate the construction of a simple molten magnesium reference electrode, for operation at temperatures greater than the melting point of magnesium (650 °C).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Long-Term Stability of Mullite and Magnesia-encased Ag|Ag + Reference Electrodes in Molten MgCl 2 -KCl-NaCl

In this study, the stability of Ag|Ag + molten chloride reference electrodes was studied using mullite and magnesia tubes as reference electrode membrane materials in molten MgCl 2 -KCl-NaCl for continuous immersion up to 31 d. Two electrochemical methods were used to characterize the long-term performance of the reference electrodes at 500 °C in an inert atmosphere using a Mg rod as the working electrode: open circuit potential (OCP) and onset of Mg(II) reduction as measured by cyclic voltammetry (CV). There was only a 3% difference between the average OCP measured versus the mullite and MgO reference electrodes. Potential drift was higher for mullite (−1.34 mV d −1 ) than MgO (0.22 mV d −1 ). But both drift rates were very low compared to the OCP data standard deviation, indicating that no significant drift was measured for us to 31 d.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Determination of Kinetic Properties of Ni(II) Ions in Molten LiF-NaF-KF via Voltammetry

Kinetic properties of Ni(II) in eutectic LiF-NaF-KF (FLiNaK) molten salt were determined at T = 748–823 K using cyclic voltammetry (CV), square wave voltammetry (SWV), and chronoamperometry (CA) measurements using a glassy C working electrode, Ni(II)/Ni reference electrode, and Ni counter electrode. Reduction of Ni(II) to Ni(s) was determined to be a single step, two-electron transfer process. Diffusivity values were calculated using the Berzins and Delahay equation and semi-integral electroanalysis from the CV measurements as well as using the Cottrell equation from the CA measurements. Diffusivity of Ni(II) in molten FLiNaK at T = 748–823 K was determined to be 3.27 × 10 –7 –3.04 × 10 –6 cm 2 s –1 with an activation energy of 62–104 kJ mol –1 . The estimated kinetic properties varied appreciably among methodologies possibly due to inherent assumptions in theory regarding reversibility and unit activity during Ni metal deposition on the working electrode.

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

Quantitative Correction of Ohmic Effects on Square Wave Voltammetry for High-Concentration Soluble-Soluble Redox Reactions in Molten Salts

Molten salts are of particular interest for a variety of industrial applications. As such, accurate characterization of species’ concentrations within the molten salt media is critical to ensure a well-controlled unit operation. Although electroanalytical tools are properly suited for precise in situ monitoring in these systems, uncompensated ohmic resistance R Ω can lead to erroneous, technique-dependent results. Using numerical simulations applied to a model system, this work will first illustrate then quantify the extent to which R Ω attenuates square wave voltammograms. This approach allows for post-experiment correction that leads to converging results across voltammetric techniques and facilitates accurate predictions of species’ concentrations.

Shaheen, Nora A. [Argonne National Laboratory (ANL↗
DOE OSTI2020

Molten Salt Reactor Experiment Benchmark Evaluation

As one of the six advanced reactor concepts selected by the 2002 Generation IV roadmap as a technology meriting future research, Molten Salt Reactors (MSRs) attracted broad attention and multiple private and public entities are working towards its commercialization. Facing stringent regulatory requirements, validations of computational codes used to calculate and prove the safety characteristics of MSRs play a key role. This project developed the world-first, MSR-related reactor physics benchmark based on a series of zero-power experiments conducted at the Molten Salt Reactor Experiment (MSRE). The benchmark was successfully reviewed by the International Reactor Physics Experiment Evaluation Project (IRPhEP) committee and is available in the IRPhEP handbook starting form the 2019 edition. To benchmark experimental data from the MSRE, a three-dimensional high-fidelity model was developed using the Monte Carlo neutron transport code Serpent 2 and ad-hoc methods have been implemented to account for the unique feature of fuel salt motion. The calculated effective multiplication factor, k eff , for the first criticality experiment, when 235 U was progressively added to the fuel salt in order to achieve criticality with stationary salt and isothermal conditions, was 1.02132 (±3 pcm). The total uncertainty for experimental k eff was estimated to be 420 pcm. The calculated k eff is 2.154% larger than the experimental and benchmark model value, which is approximately five times the benchmark model uncertainty. It is to be noted that, for systems containing large volume of graphite (or other carbonaceous materials), Monte Carlo codes tend to overestimate the keff of the benchmark model by 1% to 2%. The bias is possibly due to uncertainties in the impurity content of the graphite blocks, uncertainty in the neutron capture cross section of carbon or uncertainty in the properties of nuclear-grade graphite. The calculated reactivity coefficient of 235 U concentration (0.2228 ± 0.0014, represented as the change of reactivity over the relative change of 235 U mass in loop) matches well with the experimental value (0.223 ± 0.006), strengthening the confidence on the accurate representation of the fuel salt composition in the MSRE benchmark. Additional reactivity effect calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients show good agreement with the experimental values (within 1σ) as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Engineering scale molten salt corrosion and chemistry code development

A new engineering scale transport code Mole is described. Mole was developed using the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework for solving species transport problems in flowing liquid. It is applied to leaching of Cr from alloys into molten salts part a of a multi-physics, multi-scale approach for coupling chemistry with mechanical behavior modeling. Demonstrations of the leaching phenomenon were benchmarked to experimental data provided by the Molten Salt Reactor campaign and the results are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Evaluation of Power Fluidic Pumping Technology for Molten Salt Reactor Applications

Power Fluidics™ (PF) is a NuVision Engineering proprietary technology which uses gases to control the flow of other fluids in industrial applications. This unique technology has several advantages over traditional centrifugal pumps. Molten salt nuclear reactors and associated systems have unique fluid handling demands. This report explores the applicability of PF to molten salt systems. Several potential applications are identified, necessary modifications of existing PF technology for salt service are discussed, and a path for development and demonstration is proposed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Early Phase Molten Salt Reactor Safety Evaluation Considerations

This report documents rationale to develop analytical tools and data applicable to the safety and design of advanced molten salt power reactors. This report represents early steps in the implementation of the tasks defined by Oak Ridge National Laboratory (ORNL) in 2018 in the following sections of A Safety and Licensing Roadmap to identify the Research and Development Gaps of Commercial Molten Salt Reactors

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Predicting thermodynamic and thermophysical properties of molten chloride salts from ab-initio and classical molecular dynamics simulations

Molten salt reactors (MSRs) are among the advanced concepts pursued under the generation IV nuclear energy technology umbrella. However, the basic concept is not new and was first developed as part of the effort to power aircrafts with nuclear energy in the 1950’s. Later in the 1960’s, Oak Ridge National Laboratory (ORNL) built and operated the Molten-Salt Reactor Experiment (MSRE). This reactor used a fluoride salt with uranium as fuel. Fluorides salts are still highly relevant and proposed in several designs. In addition, chloride salts are being considered for MSRs operating in the fast neutron spectrum. This report focuses on chloride salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

FY21 Status Report on the Molten Salt Thermal Properties Database (MSTDB) Development

This report describes the thermodynamic assessments for the ZrF 4 -BeF 2 , AlCl 3 -NaCl, AlCl3-KCl and the PuCl 3 -CsCl pseudo-binary subsystems. Calculation results are compared to the inputs used to optimize the adjustable model parameters. The data came from experimental studies reported in the open literature, from the MSR research community, and/or computational results generated within the NEAMS program and from collaborators. The models from the thermodynamic assessment of these salt systems are integrated into the MSTDB-TC (Molten Salt Thermal Properties Database-Thermochemical). A brief description of the companion MSTDB-TP (Molten Salt Thermal Properties Database-Thermophysical) is given along with an overview of the approach for predicting both thermodynamic and thermophysical property behavior of multicomponent systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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

Chemical Thermodynamic Modeling of Molten Salts to Support Off-Gas Abatement Systems

The reprocessing of used nuclear fuel by any means will liberate gaseous fission products, such as hydrogen ( 3 H), carbon ( 14 C), noble gases ( 85 Kr), and halogens ( 129 I), from the irradiated fuel. These elements will distribute through chemical processing operations and partition into process off-gas streams, and the specific volatile release fractions will be dictated by the chemical and physical properties of the system. A recent assessment found that there were significant knowledge gaps regarding the release quantities of volatile radionuclides from individual unit operations. These knowledge gaps limited the ability to determine what dedicated off-gas treatment technologies could be required for electrochemical-based reprocessing facilities. Unfortunately, experimental efforts to quantify release fractions are limited by the challenges associated with performing experiments using irradiated fuel. This report documents preliminary thermodynamic predictions of iodine and tritium release from chloride-based molten salts as part of an effort to better direct resources toward those experiments (both simulant and irradiated) that will be of the greatest impact. It was predicted that less than 0.5% of tritium was expected to be released and that nearly all of that amount would be released as H 2 . Thermochemical data for hydrogen (H 2 ) are of high fidelity, and no additional validation is recommended. Less than 0.05% of iodine was predicted to be released, with the primary volatile species being Cs 2 I 2 . Unlike the tritium predictions, the data underpinning the iodine release predictions are of low quality. It is recommended that a limited experimental program be dedicated to expanding the current physical property and thermodynamic property data for iodine in electrochemical processing and molten salt conditions, including vapor pressure measurements for the dimerized species predicted to comprise the majority of iodine release. Validating and improving the thermodynamic properties used in predictive modeling can reduce the need for expensive testing with irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Considerations of Radiation-Hardened Electronics for Alpha Detectors in Molten Salt Reactors

Along with advanced generations of reactors quickly becoming reality, the need for instrumentation to provide reliable information for their safety and operation is also becoming relevant. Instrumentation-wise, molten salt reactors (MSRs) represent an especially challenging case in which detectors must survive extreme temperatures, high amounts of radiation, and a highly corrosive coolant. Adding to the complexity, tight material control and accounting must be maintained with any special nuclear material (SNM), a practice made more difficult by the nature of a loop of flowing fuel. Alpha spectroscopy offers a unique window into investigating the isotopic concentrations of actinides within just such a molten salt environment. The characteristic α emissions of specific actinide isotopes in the salt enable quantification of their concentrations. Semiconductor devices based on wide bandgap materials such as silicon carbide and gallium oxide can withstand the extreme temperatures within MSRs and offer good chemical stability with the corrosive salt. Collaborators at The Ohio State University are fabricating and testing a β gallium oxide α particle sensor that could identify actinide concentrations within the fuel salt to help improve the material control and accounting required for complicated system inherent to MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Application of SCALE to Molten Salt Fueled Reactor Physics in Support of Severe Accident Analyses

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 molten salt-fueled reactor (MSR). SCALE capabilities for the modeling of MSR physics were demonstrated based on the Molten Salt Reactor Experiment (MSRE). Of primary interest were the determination of the system’s nuclide inventory, as well as the inventories in the various regions of the loop, considering that the fuel is continuously pumped through the system. This report contains discussions on the following: 1. Determination of the system-average fuel salt inventory considering fission gas removal in the off-gas system and noble metal removal through plating out at the heat exchanger using recent enhancements in SCALE’s depletion sequence TRITON, 2. Assessment of the nuclide spatial distribution throughout the loop using SCALE’s depletion solver ORIGEN, 3. Calculation of the core’s power profile, flux profile, temperature reactivity coefficients, and xenon reactivity using full-core calculations with SCALE’s Monte Carlo code KENO-VI. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the system, as well as the inventory and decay heat of individual regions in the loop, a zone-wise power profile, temperature feedback coefficients, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Updated SAM Model for the Molten Salt Reactor Experiment (MSRE)

The development of reference standard problems based on prototypical reactor designs is of particular importance to verify the adequacy of computer codes and evaluation models for specific reactor types. To support the multiphysics coupled simulation of molten-salt-fueled reactor (MSR) using SAM and Griffin computer codes for safety and licensing analysis, much efforts have been put into enhancing code capabilities and developing reference models for the MSR primary loop in SAM. In this work, a previously developed Molten Salt Reactor Experiment (MSRE) primary loop model was updated to include a two-dimensional (2-D) core region and external core components in one-dimension (1-D) or zero-dimension (0-D). To ensure accurate feedback calculation in multi-physics simulations, the delayed neutron precursor tracking model and solid graphite model were added in the SAM model. In addition, the 2-D and 1-D domains are tightly coupled using the recently developed single-solve approach in SAM. The updated model has been tested under both steady-state and transient scenarios to demonstrate its potential for the multi-physics simulation of MSRE with coupled SAM and Griffin.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

CFD Simulations to Support Pronghorn Modeling of a Molten Salt Fast Reactor

To assist with the deployment of next generation molten salt reactors (MSRs), the DOE NEAMS program is developing advanced simulation capabilities. As part of this work, the high-fidelity CFD tool, Nek5000 was used to inform model development for the engineering scale tool, Pronghorn. The particular case chosen was the EVOL molten salt fast reactor concept. This concept uses an open core design, which is well known to be sensitive to the particular geometry and can have complex flow behavior that varies across a range of conditions. This makes it an idea candidate for demonstration of the Hi2Lo concept of using higher fidelity models to benchmark lower fidelity models. Simulations of the EVOL design were performed using both LES (high-fidelity) and RANS (moderate-fidelity) in Nek5000. The LES results indicated discrepancies in 2-D axisymmetric RANS, indicating that the core is better modeled in RANS with a 3-D wedge representation of a portion of the core, which can accurately account for the effect of the inlet channels. Finally, results from the 3-D RANS were used to modify an existing turbulence model in Pronghorn. Improvement in the calibrated Pronghorn model was demonstrated. Future work is suggested to focus on expanding the LES calculation to provide a more direct point of comparison for both the RANS and Pronghorn models as well as incorporating more of the relevant MSR physics (such as delayed neutron precursor tracking) into the model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

An Overview of the Molten Salt Thermal Properties Database–Thermophysical, Version 2.1.1 (MSTDB-TP v.2.1.1)

The current status of the Molten Salt Thermodynamic Database- Thermal Physical (MSTDB-TP) is reported. Building off of MSTDB-TP and through the direction of the Roadmap for thermal property measurement of Molten Salt Reactor systems, MSTDB-TP 2.1.1 has now been released containing a total 448 salt entries (data from 140+independent studies) compared to the original commit of 62, containing thermophysical properties including density, viscosity, thermal conductivity, and heat capacity. Along with the new database release, further advancement of Saline, a C++application programming interface (API), and a graphical user interface (GUI) has facilitated increased user/developer interaction with the database. Furthermore, estimation techniques, first principles calculations (Ab-Initio) and interpolation/extrapolation methods (Redlich-Kister/Muggianu), have shown great promise in the future of thermophysical property determination for filling out compositional spaces and investigating experimentally difficult salts (hazardous/expensive). This report describes the database composition, development and advancement of database tools, and the strategy of advancing and implementing estimation data into future iterations of the database for MSTDB-TP.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗