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DOE OSTI2021

Evaluation of Formate Salt PCM’s for Latent Heat Thermal Energy Storage

This work examines formate salts as potential phase change materials (PCMs) for middle-high temperature (≤250 °C) latent heat thermal energy storage applications. The thermophysical properties of three formate salts were characterized: pure sodium formate and binary blends of sodium/potassium formate and sodium/calcium formate. The stability of formate PCM’s was evaluated by thermal cycling using differential scanning calorimetry where sodium formate and sodium/potassium formate appeared stable over 600 cycles, while sodium/calcium formate exhibited a monotonic decrease in heat of fusion over the test period. A longer test with sodium formate led to gas release and decomposition of the salt. FTIR analysis of the PCM showed degradation of formate to oxalate. T-history experiments with 50-g PCM quantities demonstrated a bulk supercooling of only 2–3 °C for these salts. Thermal conductivity enhancement of over 700% was achieved by embedding aluminum in the solid PCM. Finally, mild carbon steel was immersed in molten sodium formate for up to 2000 h. Sodium formate was found to be non-corrosive, as calculated by mass loss and confirmed by cross-sectional high-resolution microscopy. FTIR analysis of the PCM after 2000 h shows oxidation at the free surface, while the bulk PCM remained unchanged, further indicating a need to protect the formate from atmospheric exposure when used as a PCM.phase change materials; formate salts; latent heat thermal energy storage; thermal cycling; supercooling; thermal conductivity enhancement; corrosion

25 ENERGY STORAGE↗

DOE OSTI2023

Validated CFD Model for Multimode Gasoline Compression Ignition Engine

A validated CFD model for the multimode combustion engine developed under the DOE funded project DE-EE0008478 (Co-optimized Mixed-Mode Engine and Fuel Demonstrator for Improved Fuel Economy while Meeting Emissions Requirements). It incorporates advanced physics-based fuel surrogate models for thermophysical properties and reaction kinetics. The combustion modes include spark ignition (SI), low temperature combustion (LTC), and compression ignition (CI). The real fuel model was validated for RON60, RON70, RON80, RON90, and two biofuel blends. The validation cases can be found in https://doi.org/10.2172/1887341.

02 PETROLEUM↗

DOE OSTI2020

Thermal Stability of Silica for Application in Thermal Energy Storage

Free from siting constraints, thermal energy storage (TES) shows promise as an economical alternative to traditional pumped-storage hydropower (PSH) and compressed air energy storage (CAES). As potential thermal energy storage media, many solid particles demonstrate stability over wide temperature ranges which allows for increased sensible energy storage density and is essential in achieving low-cost storage. Silica sand, in the form of a-quartz, is one such candidate. This work presents a brief review of relevant silica thermophysical properties and further investigates the thermal stability of silica particles as a candidate TES media by subjecting them to two different thermal campaigns: (1) a 500-hour thermal treatment at 1200°C under varied atmospheres; and (2) cycling 25, 50, and 100 times between 300°C and 1200°C. For both campaigns, particle stability is examined by means of pre- and post-treatment Mie scattering. An additional XRD analysis is conducted for the 500-hour treatment in air. Results indicate limited changes in both particle distribution and crystallographic structure which is promising for the application as solid particle media for thermal energy storage.

27 ARPA - Advanced Research Projects Agency-Energy↗

DOE OSTI2020

Thermal Stability of Silica for Application in Thermal Energy Storage: Preprint

27 ARPA - Advanced Research Projects Agency-Energy↗

DOE OSTI2021

Thermal Charging Rate of Composite Wax-Expanded Graphite Phase Change Materials

Phase change materials (PCMs) are valuable for their ability to store heat nearly isothermally around their phase transition temperature. PCMs are at the core of latent heat thermal energy storage (LHTES) systems, which provide the ability to buffer high thermal loads or decouple the time when heating or cooling is needed from when it is produced. Thermal charging and discharging of LHTES systems often employ a constant temperature source, and the rate at which heat can be exchanged is dependent on the thermophysical properties of the PCM. For graphite composite PCMs, the high thermal conductivity of the graphite enables an increased heat transfer rate through the material, but its presence displaces PCM which reduces the effective volumetric latent heat of the composite relative to the pure PCM. This results in a tradeoff between thermal power and volumetric energy storage capacity. The thermal charging rate is the average thermal energy stored in the material for some elapsed time. In this study, composite PCMs of compressed expanded natural graphite (CENG) and n-Octadecane are studied. Samples with varying CENG mass fractions were synthesized and the thermal charging rate was measured under a constant temperature boundary condition. For evaluation of the expanded graphite-PCM composite, one boundary of the material was exposed to a 50°C constant temperature plate, above the 27.5°C melting temperature of the PCM. The melting front progression [mm-s-1] and the thermal charging rate [W-cm-2] of the PCM were determined, and the results were compared with analytical predictions for the 1-D semi-infinite phase change. For CENG addition up to 5.75% mass fraction, a 450% thermal conductivity increase was observed with a 5% decrease in volumetric energy density as compared to pure octadecane. The average thermal charging rate was increased by over 430% for the melt to penetrate a depth of 22 mm. The experimental results matched analytical predictions, indicating that higher CENG fractions can be evaluated using analytical approaches.

Hirschey, Jason↗

DOE OSTI2021

The Effect of Working Fluid Composition on Heat Transfer in a Direct Supercritical CO2 Power Cycle Recuperator

For pure CO2, the experimental measurements of the heat transfer coefficient in this paper agreed to within 5% of the classical Dittus-Boelter correlation. When nitrogen was added to the CO2 at concentrations from 0–28% nitrogen, these experiments found that the measured Nusselt numbers decreased by 20% below the Dittus-Boelter correlation evaluated with thermophysical properties for the N2/CO2 gas mixture. In general, the McAdams correlation evaluated with mixture properties predicted the friction factor within ±10%.

Searle, Matthew↗

DOE OSTI2022

ADDITIVE MANUFACTURING OF REFRACTORY ALLOYS

Refractory alloys are known for their high temperature capabilities and are intended for use at ultra-high temperatures above even 1200°C, which exceeds the capabilities of conventional superalloys. Additive manufacturing (AM) is an attractive alternative processing route because refractory alloys are difficult to fabricate through traditional methods; AM can form a near net-shape part with a tailored microstructure. This work seeks to evaluate the solidification behavior of refractory alloys under AM conditions and establish corresponding solidification models. Two binary alloys, Mo30Nb and Nb7.5Ta, as well as the commercial Nb alloy, C103, and one refractory high entropy alloy (RHEA), MoNbTaTi, were subjected to single track melts in a laser powder bed fusion (LPBF) machine. Each melt track was evaluated in the scanning electron microscope (SEM), which determined that some AM conditions used led to non-ideal behavior. Electron backscatter diffraction (EBSD) revealed that the Nb7.5Ta, C103, and MoNbTaTi all exhibited new grain nucleation to various extents while the Mo30Nb exhibited primarily epitaxial growth. This was found to be inconsistent with solidification models that were developed for each alloy and each set of AM conditions to predict the columnar to equiaxed transition behavior (CET) of the experimental microstructures. This disparity was largely attributed to the input parameters for the solidification model, which were developed using Thermo-Calc and SYSWELD. These input parameters include the alloy specific Gibbs-Thomson coefficient and thermophysical properties, the solute specific liquidus slopes and partitioning coefficients, as well as the process specific thermal gradients. Each input parameter was evaluated to determine the likely changes required for the modeled values to generate solidification models that better correspond to what was observed experimentally.

Miklas, Abigail↗

DOE OSTI2022

Multi-Modal Characterization of Nuclear Fuels and Materials at the Idaho National Laboratory Materials & Fuels Complex

The Idaho National Laboratory (INL) leads cutting-edge research pertaining to the advancement of nuclear reactor technologies, including nuclear fuels and materials. The INL Irradiated Materials Characterization Laboratory (IMCL), Electron Microscopy Laboratory (EML), and future Sample Preparation Laboratory (SPL) are available to the nuclear research community to assess the behavior of nuclear fuels and materials, efficiently and comprehensively characterizing from the engineering to atomistic scale. The IMCL is a unique, 12,000-square-foot facility located at the INL Materials and Fuels Complex designed for analysis of irradiated materials. The facility operates advanced characterization instruments that are sensitive to vibration, temperature, and electromagnetic interference in modular radiological shielding and confinement systems, granting researchers the ability to assess the microstructural, chemical, mechanical and thermophysical properties of nuclear materials, especially irradiated fuels. The EML is dedicated to advanced characterization of materials with optical and electron microscopy tools, including scanning electron microscopy/focused-ion beam (SEM/FIB) and transmission electron microscopy (TEM). Upon construction, the SPL will be a 3 story, 49,000 sq. ft facility, that is the most modern reactor structural materials testing and analysis facility in the world, designed to investigate reactor structural materials in support of life-extension programs and development of advanced reactor concepts, including mechanical testing and advanced characterization capabilities. This presentation will showcase some of the main capabilities available at both IMCL, EML, and SPL, specifically illustrating how these characterization techniques are incorporated into multi-modal characterization work scopes to elucidate the degradation of nuclear structural materials and irradiated fuels.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Thermal Modeling and Simulation of the Packed-bed Thermal Energy Storage Combined with INL Thermal Energy Distribution System

Dynamic Energy Transport and Integration Lab (DETAIL) at Idaho National Laboratory is to support experimental demonstration and validation research on Nuclear-Renewable Hybrid Energy System [1]. The Thermal Energy Distribution System (TEDS) is a thermal-hydraulic flow loop in DETAIL with its own dedicated control system to support the integration of co-located multiple experimental systems, where a packed-bed thermal energy storage (TES) is installed as a thermal buffer and storage unit. Among various TES options, the packed-bed TES is adopted in TEDS because it offers a low-cost single-tank thermal storage option compared to the traditional two-tank TES. However, since the TES tank is filled with granular fillers having different thermophysical properties from those of TES tank wall, there is a potential thermo-mechanical issue to be carefully addressed like thermal ratcheting which may pose a significant design concern for the packed-bed TES tank. Thermal ratcheting is a thermomechanical process caused by the repeated rearrangement of granular filler inside a TES tank during continuous thermal cycling operation of the packed-bed TES system. If the thermally induced stress exceeds the yield strength of a TES tank wall, catastrophic consequences may happen like rupture of the TES tank. Thus, it is important to understand the phenomenon to ensure the robust operation of the packed-bed TES tank. Given that thermal ratcheting is caused by complex interaction of thermal transport in the porous bed and solid mechanics, the accurate prediction of transient thermal behavior of the packed-bed TES tank, which is the focus of this paper, is critical to the reliable thermal ratcheting analysis. This paper discusses the numerical modeling, simulation, and validation studies that are ongoing at INL to investigate the transient thermal behavior of the packed-bed TES. Of particular concern is the transient thermal process occurring in the packed-bed TES unit that is operated in conjunction with the INL TEDS. The main goal of this research is three-fold: (i) provide preliminary insights into the transient thermal behavior of the TEDS TES tank, (ii) support the thermal measurement and validation plan for TEDS experiment, and (iii) provide transient thermal boundary conditions to support the reliable thermal ratcheting analysis of the TEDS TES tank. For the transient thermal modeling and analysis, a CFD model was developed, and the validity of the modeling approach was examined via comparing the numerical simulation results with the experimental data obtained from various design characteristics of packed-bed TES tanks. Then, the present modeling method was applied for the transient thermal analysis of the TEDS TES tank, and the results are discussed along with the potential improvement of data acquisition strategy for the future TEDS experiments for more precise validation study.

25 ENERGY STORAGE↗

DOE OSTI2023

TMS 2023 - Methodology and Density of PuCl3-NaCl Mixtures

The development of molten salt reactors and their associated components demands reliable and accurate thermophysical property data for design and validation. A candidate fuel salt, PuCl3-NaCl (36 mol% PuCl3) has been synthesized and studied at the Idaho National Laboratory. This study has determined the density of PuCl3-NaCl at multiple compositions and temperatures, providing data for validation of modeling efforts and making improvements on the buoyancy method for density measurements of actinide-bearing molten salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

First-of-a-Kind Fuel-bearing Molten Chloride Irradiation Experiment

As a dozen MSR developers in the U.S. work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have never been irradiated. Licensing and operating these reactors requires an understanding of (1) the source term, radiation chemistry, and gas generation; (2) unanticipated irradiation-induced corrosion effects; and (3) the impact of burnup on thermophysical properties, all of which can be deduced through irradiation testing. Idaho National Laboratory (INL) is ideally suited to leverage its expertise in chloride salt chemistry, as well as its Neutron Radiography Reactor (NRAD) facility in the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC), to conduct the world’s first fuel-bearing chloride salt irradiation and fill in the knowledge gaps pertaining to salt chemistry under irradiation. Molten-salt Research Temperature-controlled Irradiation (MRTI) consists of a salt-containing capsule that is internally heated inside of a secondary containment. The experiment must rely on resistive heating to melt the salt before irradiation (to avoid the impact of radiolysis) and once the reactor is turned on and fission reactions commence in the salt, the resistive power can then be reduced. Salt-immersed thermocouples coupled to a controller allow for the heater power to be adjusted as needed to meet experimental objectives. The bulk of the experimental results will be achieved through Post-Irradiation Examination (PIE). At which point a range of different measurements are anticipated to assess the salt/plenum/wall composition, the capsule corrosion rate, and the evolution of salt properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Thermal-hydraulic Performance of Emerging Low GWP Refrigerant Mixture under Flow Boiling in Brazed Plate Heat Exchangers

The hydrofluorocarbon (HFC) refrigerants used in the current refrigeration systems are facing a phase-down due to their higher greenhouse effect resulting in global warming, and thus HVAC&R industry has undergone a transition to low Global Warming Potential (GWP) refrigerants. Refrigerant mixtures are attractive alternatives since their composition can be tailored to comply with environmental regulations while preserving favorable thermophysical properties. However, the new low-GWP zeotropic mixture refrigerants have two or more components with different saturation temperatures at the same pressure level, known as temperature glide, which can cause the degradation of the overall heat transfer performance. The brazed plate heat exchangers (BPHX) provide excellent heat transfer performance due to a compact design and are used in several air-conditioning and refrigeration applications. In this study, flow boiling heat transfer and the associated pressure drop of the refrigerant mixture in a vertical BPHX were experimentally investigated. The single-phase water-to-water experiments were conducted in the tested heat exchanger with a counter-flow configuration. The flow boiling experiments charged with R-134a and R-454C were then performed in a pumped refrigerant loop to evaluate its thermal-hydraulic performance. Furthermore, parametric studies of various heat fluxes, mass fluxes, vapor qualities, and saturation temperatures were also conducted.

Yang, Cheng-Min↗

DOE OSTI2025

Novel Synthesis and Characterization of Uranium-Zirconium Carbonitride by Direct Casting

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Integrated Molten Salt Reactor Modeling Capabilities in NEAMS Thermal Hydraulics Tools

The DOE neams program supports a full range of computational thermal fluids analysis capabilities and code developments for a broad range of advanced reactor concepts. The research and development approach under the thermal fluids technical area synergistically combines three length and time scales in a hierarchical multi-scale approach. To enable multi-scale thermal fluids capability using these codes, a key joint effort has been underway to develop an integrated system- and engineering-scale thermal fluids analysis capability, through integration of SAM and Pronghorn codes, both based on the MOOSE framework. This report summarizes recent advances in developing an integrated system- and engineering-scale modeling capability for the msr concept, which has gained significant interest in recent years. A consistent framework was established by coupling Pronghorn and SAM through the Saline interface, with thermophysical properties provided by the Molten Salt Thermal Property Database (MSTDB-TP). Further improvements were made to the coupling schemes and domain-overlapping strategies, enhancing the stability and robustness of multi-code simulations. Verification and validation efforts demonstrate the accuracy of this integration across a range of benchmark problems, including one-dimensional heated pipe flows, three-dimensional natural convection loops with evolving isotopic compositions, and \gls{msre} demonstration cases. Within Pronghorn, new capabilities were introduced to model corrosion and noble-metal plating phenomena, supported by an extended thermal-hydraulics framework and refined turbulence treatments. To capture two-phase flow behavior, a multiphase Euler–Euler model was implemented in Pronghorn, including advanced closure relations, high-resolution advection techniques, and capillary force reconstruction. Preliminary verification cases confirm the fidelity of the approach, while planned validation efforts target canonical multiphase benchmarks and application to msr components such as the msre pump bowl. Finally, updates to SAM’s msr mass transfer modeling were extended to consider noble gas migration into porous structures like graphite. The point kinetics model was updated to include reactivity feedback contributions from any defined species, such as xenon. The gas transport model was expanded for applicability to gas mixtures, bubble efflux phenomena, and species transport between liquid and gas phases. A selection of multi-scale Sherwood number correlations from MOSCATO/NekRS and multi-phase correlations from literature have been added for improved accuracy in calculating mass transfer coefficients. A companion effort on developing system-level redox corrosion has also been incorporated into SAM. Collectively, these enhancements strengthen the predictive capability of SAM and Pronghorn for simulating MSR thermal-hydraulics, corrosion, multiphase behavior, and fission-product transport, providing a more complete toolset for design, safety analysis, and licensing support of next-generation \gls{msr}s.

42 - ENGINEERING↗

DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Material Selection and Heat Transfer Model for PTUHS Device

The Path to Ultimate Heat Sink Device (PTUHS) is a passive safety system that utilizes a radiative heat transfer valve to control the rate at which heat is removed from a nuclear reactor pressure vessel (RPV) and dispersed into surrounding soil. The material selection for the PTUHS device is investigated, where the thermophysical properties are either maximized or minimized bases on what section of the PTUHS that material is being used for. Final recommendations of material choice are then presented. The heat transfer model of the PTUHS device is solved in ABAQUS, where a normal operating conditions and SCRAM conditions are solved. These temperature maps and heat flux of the system show the system's ability, where minimal heat is lost during normal operating conditions and system failure in a SCRAM scenario due to temperatures that were reached.

42 - ENGINEERING↗

DOE OSTI2025

Initial steady-state core simulation capability or thermal and pool-type molten salt reactors, coupling reactor physics, thermal-hydraulics, and evolving chemistry

This report presents the development and validation of an initial steady-state multiphysics capability for molten salt reactors (MSRs) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in Fiscal Year 2025. The framework integrates neutronics, thermal-hydraulics, species transport, and thermochemistry to capture the coupled dynamics of liquid-fueled systems. Implementation and testing were performed on two representative designs: the Molten Salt Reactor Experiment (MSRE), a thermal-spectrum, channeled-flow reactor, and the Lotus Molten Salt Reactor (L-MSR), a fast-spectrum, pool-type reactor. The modeling suite employs Griffin for reactor physics and depletion, Pronghorn and SAM for thermal-hydraulics, Thermochimica for chemistry, and Saline for thermophysical properties, with benchmarking and validation carried out against historical MSRE data, experimental flow-loop measurements, and reference depletion calculations from Monte Carlo codes. The framework demonstrated the ability to reproduce key reactor behaviors including temperature feedback, reactivity losses, delayed neutron precursor transport, xenon poisoning, and redox potential evolution. The results confirm the feasibility and accuracy of the coupled models in predicting steady-state and selected transient MSR behaviors. This latter ones are used in this report as a proxy indicating that the steady-state models from which the transient starts are accurate. For MSRE, validation showed good agreement with pump start-up and natural circulation tests, while for the L-MSR, benchmarking confirmed hydraulic calibration and consistency of neutronics–thermal coupling. The tools also provided new insights into species transport, noble metal deposition, and salt solidification dynamics. On the Xenon transport front, the code is validated against the steady state Xenon poisoining measurement and showed good agreement with the experimental value. Identified areas for future work include advanced void transport modeling, three-dimensional simulations, improved alloy corrosion models, and tighter integration with high-fidelity Monte Carlo codes. These developments provide a foundation for high-fidelity MSR simulations that can support reactor design optimization, safety assessments, and long-term operational strategies.

42 - ENGINEERING↗

DOE OSTI2025

On Uranium-Zirconium Carbonitride Formation by Direct Casting - A Novel Synthesis Study

36 - MATERIALS SCIENCE↗