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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

System Design of a 2.0 MWth Sodium/Molten Salt Pilot System

Nitrate molten salt concentrating solar power (CSP) systems are currently deployed globally and are considered state-of the art heat transfer fluids (HTFs) for present day high-temperature operation. Although slightly higher limits may be possible with molten salt, to fully realize SunShot efficiency goals of $15/kWhth HTFs and an LCOE of 6¢/kWh, HTF technologies working at higher temperatures (e.g., 650 °C to 750 °C) will require an alternative to molten salts, such as with alkali metal systems. This investigation explores the development of a 2.0 MWth sodium receiver system that employs a sodium receiver as the HTF, as well as with a ternary chloride (20%NaCl/40%MgCl/40%KCl by mol wt.%) salt as a thermal energy storage (TES) medium to facilitate a 6-hr. storage duration. A sodium-to-salt heat exchanger model as well as a salt-to-sCO2 primary heat exchanger model are employed and evaluated in this investigation. A thermodynamic system design model was developed using Engineering Equation Solver (EES) where state properties were calculated at inlets and outlets along both hot and cold legs of the pilot-scale plant. This investigation assesses receiver performance as well as system efficiency studies for the pump and system operational ranges. Results found that high efficiency sodium receivers were found to have higher heat transfer coefficients and required far less spreading of incident flux. The system performance model results suggest that for a pump speed of 2400 RPM, respective hot and cold pump TDH values were determined to be 260.1–307 ft. and 260.1–307 ft for pump flow rates of 90–120 GPM.

41 EE - Solar Energy Technologies Office (EE-4S)↗
DOE OSTI2026

Processing iridium foil targets for the production of radioplatinum isotopes

Dissolving noble metals remains challenging because of their high reduction potentials and tendency to passivate as oxides. This barrier is acute for iridium metal targets used to produce platinum radioisotopes, where post-irradiation processing is costly and hazardous with conventional routes. Here, we report a rapid alkali-fusion method that converts Ir metal directly into chloride media compatible with downstream separations. In the process, an Ir foil is fused with Na 2 O 2 and KCl at temperatures above the KCl melting point; the molten salt enhances heat transfer and facilitates O 2− /Na + transport to the Ir surface, overcoming Ir's oxidation resistance. The fused cake is then cooled and leached in 10 M HCl, yielding Ir in solution without specialized microwave digestion or HF. Pt separation from bulk Ir was evaluated using extraction-chromatography resins. A newly formulated ionic-liquid resin containing Cyphos IL 104 (E104) was compared with two commercial resins (TrisKem TK200 and TK201). All resins achieved >90 % Ir recovery. Pt co-recovery depended on resin chemistry: TK200 recovered 85.9 % Pt, TK201 recovered 54.1 % Pt, and E104 recovered 72.9 % Pt. This study introduces a fast, scalable dissolution route for solid Ir targets and demonstrates resin choices that tune Ir/Pt separation, lowering the barrier to routine production and purification of Pt radioisotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

MSTEC: Molten Salt Thermophysical Examination Capability

Technical Description – a shielded modular hotcell with an inert argon atmosphere, housing characterization equipment for determining thermophysical and thermochemical properties of high temperature liquids not limited to but focusing on irradiated fuel salts Location – MFC, FCF, rm 35 Compatible Materials: Chloride, fluoride salts Fresh fuel salts and irradiated fuel salts Pyrophoric material - U, Pu metal Gases – H2, HCl, Cl2, HF, F2, NF3 Beryllium containing salts Many others

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Electrochemical Deposition with Redox Replacement of Lanthanum with Uranium in Molten LiCl-KCl

Electrochemical recovery of dilute concentrations of actinides from spent nuclear fuel would reduce the longevity of storing high-level nuclear waste. Electrochemical deposition with redox replacement (EDRR) is used in a molten salt medium for the selective electrochemical recovery of uranium in the presence of excess concentrations of lanthanum. In each EDRR cycle, after a short electrodeposition pulse, the deposited lanthanum is spontaneously replaced by uranium at open circuit. After repeated cycles, uranium metal was obtained on a tungsten electrode immersed in LiCl-KCl melt that contained 1 wt.% lanthanum chloride – 0.15 wt.% uranium 4+ chloride. Here, scanning electron microscopy and energy dispersive X-ray spectroscopy analysis revealed uranium particles approximately 0.5 - 1 μm with well-defined rectangular shapes; and with 20 – 60 times more uranium recovered on the surface of the electrode than lanthanum.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Multi-Physics Simulations of Molten Chloride Fast Reactor using Nek5000 and PROTEUS-NODAL

This report documents the FY20 work to demonstrate a specific multi-physics capability within NEAMS tools to model molten chloride fast reactors (MCFR). Specifically, the CFD code Nek5000 and the neutronics code PROTEUS-NODAL were coupled off-line to calculate the delayed neutron precursor distribution within the flowing fuel salts in an MCFR core. An MCFR benchmark problem and a 3D test problem based on a conceptual MSR design were successfully solved through the coupled simulation of Nek5000 and PROTEUS-NODAL. In the coupled simulations for these test problems, Nek5000 provided high-quality flow field information, and PROTEUS-NODAL performed the neutron diffusion calculation along with the precursor drift model. The impact of precursor drift on the core status was also investigated. Additionally, the precursor drift model implemented in PROTEUS-NODAL was verified by making use of the reference precursor distribution generated from Nek5000. Results of these test problems suggest that the multi-physics capability of Nek5000 and PROTEUS-NODAL was properly developed for tracking delayed neutron precursor drift.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Design and Fabrication of a Temperature-Controlled Fueled Molten Salt Capsule Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

ENYGF 2023 - Design of Temperature-controlled Fuel-salt Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Kinetic Parameters and Diffusivity of Uranium in FLiNaK and ClLiK

Anodic stripping voltammetry (ASV) and cyclic voltammetry (CV) measurements at 773, 823, and 873 K were made of uranium trifluoride (UF 3 ) in lithium fluoride-sodium fluoride-potassium fluoride eutectic (FLiNaK) and uranium trichloride (UCl 3 ) in lithium chloride-potassium chloride eutectic (ClLiK). ASV data were used to estimate the charge transfer coefficients, exchange current densities, and activation energies of the uranium reactions. Charge transfer coefficients of both salt systems were within the range of 0.17 to 0.38. Exchange current densities in the fluoride and chloride salts were estimated within the range of 0.060 to 0.12 A cm -2 . Activation energy of uranium exchange current was 16.0 kJ mol -1 in the fluoride salt and 28.9 kJ mol -1 in the chloride salt. Kinetics of charge transfer were found to be faster in FLiNaK. Analyses of the CV data suggest the electrochemical system was diffusion controlled and irreversible in the chloride and fluoride salt mixtures. Diffusion coefficients of uranium in the range of temperatures were on the order of 10-5 cm 2 s -1 in both systems. Greater values of diffusivity in ClLiK are attributed to its lower density compared with FLiNaK. Activation energy of uranium diffusion in the fluoride and chloride salt mixtures were 53.4 and 89.4 kJ mol -1 , respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Characterization of MC&A for the Molten Salt Fuel Cycle

Advanced reactor developers are exploring diverse reactor designs, including molten salt reactors (MSRs). These advanced reactors are considered for wider applications and a range of deployment locations, including supporting the integration of renewable energy sources in the grid. There are three main types of MSRs: (1) reactors in which the fuel salt freely circulates within the core; (2) reactors with the fuel salt contained within vented fuel tubes; and (3) reactors that use molten salt solely as a coolant, with the fuel in a separate, solid form. In this document, the term MSR refers specifically to the first two types, which use fuel salt—special nuclear material (enriched uranium, plutonium, and 233 U) in chloride or fluoride form mixed with chloride- or fluoride-based carrier salt in a peritectic mixture—as the primary medium for fission. The composition of fuel salt, both at startup and for makeup or refueling, varies depending on the MSR design and the chosen fuel cycle approach, which can be either once-through or closed. For MSRs, a variety of fuel cycle approaches (e.g., U, U–Pu, U–Pu–TRU, U–Th, U–Pu–Th) are being considered. Fuel in MSRs is much different than traditional solid fuel, including its preparation. The uniqueness warrants investigation into characterizing fuel preparation processes, known as fuel salt synthesis . This effort characterized major fuel preparation and synthesis processes, identifying temperature, equipment, and environmental requirements for uranium-, plutonium-, and thorium-based fuel preparation and synthesis. Because MSR fuel salt synthesis facilities handle special nuclear material in loose, bulk form, a material control and accounting plan will be required for licensing from the US Nuclear Regulatory Commission or under the US Department of Energy authorization. This effort serves as a foundation to investigate material control and accounting approaches for synthesis facilities, including determining measurement points and techniques. Because several MSR developers are planning demonstration facilities in the coming years, this effort will support stakeholders with preparing or reviewing material control and accounting plans for providing assurance that all special nuclear material is accounted for at fuel salt synthesis facilities. This report was produced for Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy (DOE), Office of Nuclear Energy, Nuclear Fuel Cycle and Supply Chain.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Heterogeneous 3D Morphological Evolution of Ni Microparticles in Molten Salts: Visualized by Operando Synchrotron X-ray Nano-tomography

Ni-based superalloys are promising materials for high-temperature molten salt (MS) energy generation and storage. Studying morphological and chemical evolution of pure Ni in MS provides fundamental knowledge for MS technologies and corrosion mitigation. Here, in this study, real-time 3D morphological changes of Ni microparticles in molten KCl-MgCl 2 were studied by operando synchrotron X-ray nano-tomography at 700°C. Rapid Ni particle agglomeration occurred, without significant chemical reactions, such as oxide or chloride formation. The morphological growth evolved differently from classical coarsening or sintering behaviors and occurred nonuniformly, with other regions showing slight dissolution of Ni. Ni nanoparticles were found to be dispersed in many areas of the samples, either from microparticle dissolution or other radiation-induced nanoparticle formation mechanisms. This study discusses important factors, i.e., thermal gradient, amounts of salt and metals, and radiation effect, that influence morphological changes of materials in MS, critical for fundamental understanding of material–MS interactions as well as for practical applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Special Considerations for the Removal and Disposal of Micro-Reactor Experiments

Idaho National Laboratory (INL) is preparing to host several microreactor experiments through 2030 and beyond. Two new test beds currently under development will operate as microreactor or nuclear system experiment user facilities. Test bed experiments will be performed in series, with each nuclear experiment installed, operated, and removed before installation of the subsequent experiment. The necessarily brief transition period between experiments introduces unique equipment removal and radioactive waste disposal challenges. This paper evaluates equipment removal and radioactive waste management topics associated with tight sequencing of nuclear experiments and presents some of the solutions and approaches currently planned to meet these special considerations for one such experiment, the Molten Chloride Reactor Experiment (MCRE), slated for operation at INL’s Laboratory for Operation and Testing in the United States (LOTUS). The MCRE project is a collaboration between Southern Company Services, TerraPower, and INL, among others, to provide integral nuclear data that will advance molten salt fast reactor technology. MCRE will be the first critical fast-spectrum circulating fuel system ever operated and the first experiment operated in the LOTUS test bed. The experiment will nominally operate at zero power with planned low power excursions as part of operational planning for MCRE has been to minimize at-power operations to limit fission product formation while still achieving experimental objectives. After the experiment is complete, MCRE will be allowed to radioactively decay for a short period (nominally 90 days) prior to system defueling, flushing, removal, and disposal of all MCRE equipment, readying the test bed for the next nuclear experiment. Equipment removal and radioactive waste disposal have been integral to MCRE project planning since the Cooperative Research and Development Agreement was formalized in 2021. Due to requirements for future use of the test bed, the MCRE system, of necessity, must be removed in a much shorter time frame than typical for historic reactor decommissioning projects at INL. This results in minimal time for radioactive decay, resulting in not only elevated radiation dose rates but also the presence of short-to-medium-lived isotopes not typically encountered in the reactor decommissioning and radioactive waste management space. Additional unique constraints placed on the project include lack of intrinsic remote-operations capabilities in the test bed, space constraints in the test bed once MCRE has been installed, and contamination minimization requirements to return the test bed to as-found conditions to enable future use. This paper discusses planned solutions to these challenges. Approaches for implementing remote or semi-remote technologies in a non-hot cell environment with limited space availability are discussed. The paper also summarizes the systems engineering approach for concept development and design of equipment removal systems, which are being implemented concurrent with the MCRE design phase, providing feedback to system designers to incorporate features enabling efficient and safe equipment removal approaches. The timing of this paper at a relatively early phase of the project is meant to highlight the importance of early planning for nuclear system decommissioning while reactor design is ongoing to allow design feedback on componentry driven from decommissioning system needs. As new, innovative nuclear reactor technologies enter the nuclear market sector, reactor experiments are crucial for providing new integral nuclear data that establish safe operational margins for technology advancement. Microreactor experiments at INL will require safe, effective, and timely decommissioning approaches, which in turn require nuclear systems designed to expedite decommissioning. In addition to the advancement of nuclear reactor technologies, these nuclear experiments provide an opportunity to demonstrate, deploy, and test new removal and disposal capabilities to support the next generation of nuclear reactor technology.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

In-Situ TEM Molten Salt Corrosion

Molten salt reactors (MSRs) offer a compelling pathway for next-generation nuclear energy, with advantages in thermal efficiency, inherent safety, and flexible fuel management. Yet, halide-based molten salts introduce significant materials challenges, particularly alloy corrosion. Alloy performance in these environments ultimately depends on understanding how corrosion initiates and progresses at the nanoscale, however most existing models rely on post-exposure characterization, leaving degradation mechanisms largely inferred rather than directly observed. NiCr alloys have garnered interest in MSRs applications as the Ni-based matrix provides strength and creep resistance, while Cr content offers oxidation resistance in air. However, NiCr corrosion resistance in chloride salts has proven poor due to preferential chromium dissolution, the formation of Cr-depleted pathways, and grain-boundary attack. This work aims to directly visualize corrosion of Ni-20Cr exposed to LiCl-KCl using in-situ Transmission Electron Microscopy (TEM) to capture real-time microstructural evolution during corrosion. Experiments will be performed at ~800 °C under controlled pressure conditions while utilizing Energy-Dispersive X-ray Spectroscopy (EDS) to analyze elemental redistribution. Observation of chromium depletion fronts, associated surface restructuring, and localized chloride enrichment are expected. Ultimately, this study is expected to provide a link between microscale processes and the macroscopic degradation behaviors relevant to MSR operation in advanced reactor environments. Simultaneously, this approach enables future in-situ investigations regarding alloy composition, salt chemistry, and their influence on corrosion pathways and long-term stability.?

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Electrochemical measurement and analysis of YCl 3 , ScCl 3 , GdCl 3 and MgCl 2 in molten eutectic LiCl-KCl

For this work, electrochemical measurements of YCl 3 , ScCl 3 , GdCl 3 and MgCl 2 were obtained in molten LiCl-KCl eutectic at 773 K using a three-electrode configuration. Tungsten rods were used as the working and counter electrodes. Two quasi- and one reference electrode(s) (glassy carbon, nickel wire, and Ni/NiO) were used to collect electrochemical measurements of the analyte salts. Cyclic voltammetry data were analyzed to calculate values of diffusion coefficient, exchange current density, and charge transfer resistance and were determined to be on the order of 10 –5 cm 2 s –1 , 10 –2 A cm –2 , and 10 0 Ω; respectively. In general, the values of diffusion coefficients were found to be consistent with those reported of high-temperature molten salts. Relatively large values of exchange current density corresponded with smaller values of charge transfer resistance. These values were found to be reasonable in comparison to results available in the literature. The measured cyclic voltammograms were normalized with respect to both lithium reduction and chloride ion oxidation potentials. Such a normalization technique is effective for comparing experimentally obtained cyclic voltammetry data to those that have been published. An analysis of experimentally obtained results indicates the relation of electrode reactions (e.g., differences of reduction potentials) are independent of the choice of reference electrode. Additionally, the choice of reference electrode did not affect the electrochemical window, exchange current density, or diffusion coefficient values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Exploring the speciation of actinide salts in the presence of contaminants related to molten salt reactors

Molten salt reactors (MSR) are up-and-coming Generation IV nuclear reactors with either coolant and/or radioactive fuel in molten salt form. Due to their improved safety, efficiency, affordability and convenient waste processing system, these reactors are considered a superior alternative to conventional ones. However, prior to commercial implementation, it is necessary to have an extensive understanding of the reactions occurring in the MSR. The safety parameters of the MSR are determined by the rheological properties such as density and viscosity of the molten salt and these properties are governed by the local structure of the molten salts. Due to the high operational temperatures of MSR's, corrosion plays an important role. Because of their influence, the container corrosion products become part of the fuel salt matrix. In this work, we are investigating how these corrosion products, e.g., Mo, Ni, Cr, would impact the speciation of the actinide metal centers. For the initial studies, lanthanides are used as a surrogate for actinides. Lanthanide chloride salts (LnCl3) are mixed with transition metals and excess of alkali or alkali earth metal salts or salt mixtures (LiCl, NaCl, KCl). These are added to an alumina crucible, placed in a quartz tube, and sealed under a vacuum. This is heated up to around 1000 °C and slowly cooled to a temperature above the melting point of the salt and at that temperature, the reaction is taken out of the furnace. The excess salt is removed, and the resulting product is analyzed using X-ray diffraction techniques, scanning electron microscopy, and spectroscopic methods such as Raman, UV-Visible, and FTIR spectroscopy.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2025

Benchtop-Scale High Temperature Molten Salt Electrochemical Reactors: Experimental Setup and Considerations

Molten salt electrochemical reduction (MSER) is the dominant production method for aluminum and titanium, and emerging technologies for producing other commodity metals with these highly electrified techniques have demonstrated promising results. However, the scale up of MSER of materials such as iron and silicon are limited by the challenging issues of materials compatibility and impurity control. This work describes an experimental MSER setup using both chloride and carbonate salts. This setup is designed for operations outside of a glovebox, which is practical from an industrial perspective, but creates unique challenges. This work serves as a practical guide to discuss some key operational difficulties such as temperature control, air and humidity, material compatibility, developing strong electrical measurements and techniques, and safety.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗