Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Steam Generator”

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.

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

At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Passive cooling to cold shutdown

A cooling system for a reactor module includes a reactor pressure vessel that houses primary coolant and a steam generator that lowers a temperature of the reactor pressure vessel by transferring heat from the primary coolant to a secondary coolant. The steam generator releases at least a portion of the secondary coolant as steam. Additionally, the cooling system includes a containment vessel that at least partially surrounds the reactor vessel in a containment region. The containment region is dry during normal operation of the reactor module. A controller introduces a source of water into the containment region in response to a non-emergency shut down of the reactor module. The source of water is located external to the containment vessel, and the water is introduced into the containment region after the steam generator has initially lowered the temperature of the reactor pressure vessel in response to releasing the steam.

Keller, Michael↗
DOE OSTI2016

Effects of bypass flow and power change and deviation on performance of thermal mixing structure of HTR-PM - 18424

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power because of power regulation will introduce higher radial temperature difference of coolant helium from reactor to steam generator. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the mixing performance of the thermal mixing structure at HTR-PM reactor core outlet. The experiments on Test Facility Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation on mixing performance. In addition, some simulation calculations are conducted for the main mixing structure of the TF-HGM facility and the thermal mixing structure of HTR-PM. According to the results of experiments and simulation calculations, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, less than 30 deg. C ({+-}15 deg. C), under all considered conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Effects of wall thickness on the flow instability of supercritical pressure water in a tube - 18568

The subcritical steam generators in the current high-temperature-reactor pebble-bed module, HTR-PM, in China can be replaced by supercritical steam generators to better coordinate the reactors and the supercritical steam turbines to improve thermal efficiencies with no phase change at supercritical pressures. This study numerically studied the heat transfer and flow instability of supercritical pressure water in a vertical tube with zero- and finite-thickness walls. The effect of the wall thickness on the heat transfer of supercritical pressure water at steady state was negligible. In the transient calculations, the fluid flow rate oscillated intensely beyond a certain heat flux with zero-thickness wall, while the flows in the tube with the finite-thickness walls acted totally different. The wall properties including the density and specific heat were varied to analyze why the oscillations were suppressed when the finite-thickness wall was considered in the calculation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1986

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Kadambi, J. R.↗
DOE OSTI2017

Control system design for a small pressurized water reactor - 141

Small Pressurized Water Reactors (SPWR) have unique designs from the commercial large Pressurized Water Reactors (PWRs). Integral design is adopted. There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. Strong interactions exist among different loops. The pressure vessel is large in volume and the coolant inventory is also in a large amount. The inertia of the coolant is large and it takes more time for the primary system to respond to disturbances. The steam generator adopts the once-through design and the water inventory is small. The steam generator is very sensitive to disturbances. These unique characteristics challenge the control system design of SPWR. Relap5 is used to model the SPWR. Steady-state simulation is performed to verify the correctness of the model. In the reactor power control system, both the reactor fission power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance for the control system. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow and the letdown flow is compensated to calculate the charging flow needed. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Coiled Tube Gas Heaters For Nuclear Gas-Brayton Power Conversion

This project developed an alternative design for heat exchangers for application to heating supercritical carbon dioxide (S-CO 2 ) or air for power conversion. We have identified an annular coiled tube bundle configuration–where hot sodium enters tubes from multiple vertical inlet manifold pipes, flows in a spiral pattern radially inward and downward, and then exits into an equal number of vertical outlet manifold pipes–as a potentially attractive option. The S-CO 2 gas or air flows radially outward through the tube bundle. Coiled tube gas heaters (CTGHs) are expected to have excellent thermal shock, long-term thermal creep, in-service inspection, and reparability characteristics, compared to alternative options. CTGHs have significant commonality with modern nuclear steam generators. Extensive experience exists with the design, manufacture, operation, in-service inspection and maintenance of nuclear steam generators. The U.S. Nuclear Regulatory Commission also has extensive experience with regulatory guidance documented in NUREG 0800. CTGHs leverage this experience and manufacturing capability. The most important difference between steam generators and gas-Brayton cycles such as the S-CO 2 cycle is that the heat exchangers must operate with counter flow with high effectiveness to minimize the pinch-point temperature difference between the hot liquid coolant and the heated gas. S-CO 2 -cycle gas heaters also operate at sufficiently elevated temperatures that time dependent creep is important and allowable stresses are relatively low. Designing heat exchangers to operate in this regime requires configurations that minimize stresses and stress concentrations. The cylindrical tubes and cylindrical manifold pipes used in CTGHs are particularly effective geometries. The first major goal of this research project was to develop and experimentally validate a detailed, 3-D multi-phase (gas-solid-liquid) heat transport model for CTGHs, using methods similar to earlier UCB multi-scale models for PCHEs, which will enable optimization of CTGH designs with respect to the number of manifold pipes, S-CO 2 and sodium circulating power, and other design parameters. CTGHs are relatively novel because the tubes loaded in compression as well as the manifold pipes which form the tube sheets. The second major goal of the experimental and modeling effort was to optimize the assembly and heat treatment of the tube-to-tubesheet joints, using a novel tapered joint configuration, and to develop diffusion-bonded joints with high shear strength and creep resistance, as has been done in other applications for tapered plugs. The project would also conduct creep testing of these tubesheet joints in pressurized furnaces under prototypical conditions of stress and temperature, however this was not feasible due to numerous hinderances. Detailed microstructural investigations was conducted on the diffusion bond interface, while the bond strength was evaluated by tensile testing. The mechanical testing quantified the overall strength of the bond and measured the pull out force as well as long term behavior, which complemented the microstructural analysis and gave a complete understanding of the diffusion bond.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2024

Flow Instabilities in boiling channels and their suppression methodologies—A review

Small modular reactors (SMRs) are gaining significant attention as a promising solution for clean and sustainable nuclear-power generation. However, the operation of SMRs is subject to various challenges, including two-phase flow instabilities. Flow instability has the potential to trigger flow-induced vibration and cyclic fluctuations in local thermal stress. These instabilities frequently manifest because of the complex interplay among a multitude of factors, encompassing thermal-hydraulic conditions, the geometric configuration of the steam generator, and operational parameters. These conditions could subsequently lead to premature critical heat flux, equipment malfunctions, and other safety concerns. Here, the endeavor to address steam-generator flow instabilities is of utmost importance in augmenting the sustainability and efficiency of contemporary energy production. This study offers a comprehensive review of instabilities in two-phase flow, with a particular focus on the influential factors impacting the stability of flow boiling. Furthermore, it delves into the processes of identifying, characterizing, and ameliorating these instabilities, emphasizing pivotal findings, methodologies employed, and avenues for prospective research. The primary parameters of concern encompass the efficient transfer of thermal energy, the optimization of mass-flow rates, and the establishment of favorable boundary conditions, all in the context of steam generator design to alleviate instability for water-cooled SMRs. These discernments bear substantial ramifications for enhancing reactor performance and ensuring operational safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Innovative Decommissioning Strategies for Large Components - 18535

Within decommissioning projects the dismantling of steam generators as well as of the reactor pressure vessel with its internals form important and time consuming tasks. In order to optimize these, the reactor pressure vessel will be segmented into larger pieces in-situ and the larger pieces will be further segmented in a post segmentation area in order to suit the dimensional and radiological requirements of the final waste package. A similar approach is used also for the dismantling of Steam Generators. NUKEM Technologies Engineering Services has combined the two principles 'Rip / Store / Ship' and 'In-situ segmentation' to firstly segment the steam generator in-situ and then pack and ship all resulting parts allowing decontamination within 20' container to an existing decontamination facility. The aim is to free release 90 - 95% of the material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

SAM Two-Phase Flow Model Development and Applications for Operational Transients in Advanced Reactors

As advanced nuclear technologies continue to develop, the need for the flexible operation and generation of these advanced reactors becomes necessary to maximize economic potential. As large-scale experiments are not always feasible, modeling and simulations of advanced reactors play a crucial role in design optimization and analysis. The SAM (System Analysis Module) code developed at Argonne National Laboratory is a state-of-the-art system-level thermal-hydraulic code aimed at simulating advanced reactor systems. Recent code developments have implemented two-phase flow modeling using the homogeneous equilibrium model, and a new steam generator component has been developed to utilize the two-phase flow implementation. In addition to verification tests, a load-following simulation was performed to model a realistic load-following transient in a proposed integrated system consisting of a conceptual advanced reactor known as the Advanced Burner Test Reactor (ABTR) and thermal energy storage (TES) tanks. The integrated system model uses two large TES tanks designed for sodium and a model helical coil steam generator to simulate the operational load-following transient. The flow rates of the feedwater and secondary loops are regulated to meet a prescribed steam generator load consistent with the electricity demand over a 24-h period. In conclusion, the results found the ABTR system was able to maintain stable reactor conditions and primary- and secondary-side characteristics over the course of the load-following transient.

Advanced Burner Test Reactor (ABTR)↗
DOE OSTI2015

Eddy current signal comparison for tube identification

Inspection of nuclear power plant steam generator tubes is required to justify continued safe plant operation. The steam generators consist of thousands of tubes with nominal diameters of 15 to 22mm, approximately 1mm wall thickness, and 20 to 30m in length. The tubes are inspected by passing an eddy current probe through the tubes from tube end to tube end. It is critical to know exactly which tube identification (row and column) is associated with each tube's data. This is controlled by a precision manipulator that provides the tube ID to the eddy current system. Historically there have been some instances where the manipulator incorrectly reported the tube ID. This can have serious consequences including lack of inspection of a tube, or if a pluggable indication is detected, the tube is likely to be mis-plugged thereby risking a primary to secondary leak.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

Optimal design of phase change material storage for steam production using annual simulation

A direct steam generation (DSG) solar collector can be an efficient process-heat application compared to electricity generation, and it can significantly reduce dependence on non-renewable energy sources for steam generation. Due to intermittency of the sun, thermal energy storage is required to increase the capacity factor or the yearly utilization of the system. For this study, phase change material (PCM) is used to store the heat from the steam produced in the solar collector. Sodium formate is selected as the PCM material on a cost basis. A detailed off-design analysis is carried to predict the annual performance of the system. Fins are used to enhance the thermal performance of the PCM, and an simplified method is derived to predict the required fin mass. With an increase in the number of fins, the thermal performance of the system increases significantly at design point. Furthermore the fins are expensive, so an annual simulation and techno-economic analysis is carried to calculate the optimal number of fins and other operating parameters.

14 SOLAR ENERGY↗
DOE OSTI2017

Operator Action Time Evaluation of Emergency Water Injection Strategy for Extreme SGTR Accidents

In terms of severe accidents caused by earthquakes or tsunamis that are beyond expectation, a special safety inspection for operating plants, following the Fukushima accident, has been conducted by the government of Korea to verify that nation's nuclear power plants have been adequately designed to respond to extreme accidents. The inspection results show that Korean nuclear power plants in operation have no imminent risk for an expected maximum potential earthquake or coastal flooding based on the up-to-date investigation. However, there is a need to implement long- and short-term safety improvements in order to secure the level of safety for natural beyond-design-basis events. One of the measures to increase the mitigation capability during extreme accidents is installing the injection flow paths to provide emergency cooling water of extemal sources using fire engines to the steam generators (SGs) or reactor cooling system (RCS). An emergency cooling water system consists of a fixed pipe connected from the RCS or SGs to the outside of the containment. A standby valve is installed on the pipe. Following the occurrence of an extreme accident, movable equipment (for example, a fire truck hose) can be connected to the pipe hole at the opening of the isolation valve. The emergency cooling water injection strategy was not taken into account for the steam generator tube rupture (SGTR) accident management strategy of the OPR-1000. Therefore, it is necessary to develop some guidelines or strategies to cope with an extreme SGTR scenarios using the newly installed injection flow paths and fire engines. As long as the fission products released from the core during severe accidents are kept inside the containment, there may be several ways to reduce the radionuclide release into the environment, for example, by using plant safety features. The containment bypass sequences, however, allow the fission products to be released into the environment or directly to the auxiliary building. The SGTR accidents, which are one of the containment bypass sequences, have shown during various probabilistic safety assessments (PSAs) to be contributors to plant risk, mainly because of their potential for causing a release outside containment. This paper is intended to be an investigation of the characteristics of accident progressions and assessments for the mitigative effectiveness of the external cooling water injection strategies during extreme SGTR accidents in a typical OPR-1000 plant. Analyses were performed using a MAAP computer code version 5.03. The MAAP code is a system-level computer code capable of conducting integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 PSA or severe accident management strategy developments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1979

Phase 1 of the First Small Power System Experiment (engineering Experiment No. 1). Volume 3: Experimental System Descriptions

The design and development of a modular solar thermal power system for application in the 1 to 10 MWe range is described. The system consists of five subsystems: the collector, power conversion, energy transport, energy storage, and the plant control subsystem. The collector subsystem consists of concentrator, receiver, and tower assemblies. The energy transport subsystem uses a mixture of salts with a low melting temperature to transport thermal energy. A steam generator drives a steam Rankine cycle turbine which drives an electrical generator to produce electricity. Thermal and stress analysis tests are performed on each subsystem in order to determine the operational reliability, the minimum risk of failure, and the maintenance and repair characteristics.

Holl, R. J.↗
DOE OSTI2022

Front-End Engineering Design Study for Retrofit Post-Combustion Carbon Capture on a Natural Gas Combined Cycle Power Plant

The objective of the project is to conduct a Front-End Engineering Design (FEED) study to determine the technical and economic feasibility of installing a retrofit, post-combustion, carbon capture facility on a commercially operating, natural gas-fired, combined cycle (NGCC) power plant. The Electric Power Research Institute (EPRI), California Resources Corporation (CRC), and Fluor Corporation used Fluor's Econamine FG Plus SM (EFG+) conducted the FEED study for capturing CO 2 produced by CRC's 550 MWe Elk Hills Power Plant (EHPP), located in the Elk Hills Oil Field near Tupman, Kern County, California. The EHPP was commissioned in 2003 and is powered by two General Electric 7FA gas turbines, with two heat recovery steam generators (HRSGs) providing steam to a General Electric D11 steam turbine. The target capture amount is 4,000 tonnes CO 2 /day for use in either enhanced oil recovery or dedicated geological saline storage located on CRC property at or nearby EHPP. This CO 2 is captured from a combination of the CO 2 emitted from the flue gas from EHPP and the flue gas generated from a natural gas-fired auxiliary boiler that supplies steam to the EFG+ process.

03 NATURAL GAS↗
DOE OSTI2022

Electrification of industrial boilers in the USA: potentials, challenges, and policy implications

Decarbonization of the industrial heat demand through electrification could contribute significantly to climate change mitigation efforts. In the US industry, thermal processes accounted for 75% of the total final energy demand in 2018, of which 17% is consumed by conventional industrial boilers (excluding cogeneration) for steam generation. Electric boilers have a small share in the US industrial steam generation due to several techno-economic reasons. Here, this study employs a bottom-up approach to investigate the sector-level and state-level techno-enviro-economic potentials of deploying electric boilers in the US industry in different timeframes. The results show that the technical potential energy and CO 2 savings by electrifying industrial boilers are 595 PJ and 200 MtCO 2 per year in 2050, respectively; however, these incur additional costs in each sector. Although there may be individual cost-effective opportunities for electrifying boilers in specific industrial sites, the overall costs are high in all industrial sectors and states due to the large disparity between electricity and combustion fuel prices. To overcome the highlighted techno-economic barriers, a comprehensive action plan for different stakeholders is also formulated. This study provides novel insights that should inform policymakers’ and executives’ decisions about the electrification of the current and future US industrial boiler systems.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2015

Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

United States Nuclear Manufacturing Infrastructure Assessment

This report provides a review and assessment of the existing U.S. infrastructure capabilities to manufacture on a commercial scale, the major components and unique materials required by the various U.S. designs for advanced nuclear reactors and small modular reactors (SMRs). The approach was to compare the needs of SMRs and advanced reactors against the capability of U.S. manufacturers to ascertain gaps that could inhibit U.S. industrial support for deployment of SMRs or advanced reactors. This report focused on the reactor plant designs being developed by NuScale Power (the NuScale Power Module), TerraPower (both the Traveling Wave Reactor and the Molten Chloride Fast Reactor to a limited extent), and X-energy (the Xe-100). To gather the information, MPR visited each of these companies and selected candidate suppliers to tour their facilities and conduct in-person interviews. Information was also obtained by telephone interviews, internet research, and literature review of technical reports. The overarching conclusion from this study is that the U.S. has existing capability or can readily develop capability to support and sustain most, but not all, aspects of commercial deployment of advanced reactors and SMRs. The greatest concerns for capability gaps in the U.S. infrastructure for supply of components for SMRs and advanced reactors are associated with reactor pressure vessels, steam generators, and fuel fabrication. Foreign vendors can supply reactor pressure vessels and steam generators. Issues with fuel fabrication are associated with uranium enrichment and development of fabrication methods for the innovative fuel concepts planned for advanced reactors. The technical capabilities of U.S. manufacturers appear satisfactory with the possible exceptions noted above. However, the capacity of the existing U.S. manufacturing infrastructure may be insufficient for a high rate of construction. Manufacturers contacted as part of this study indicated a willingness to expand capacity if sufficient orders are in place. The review of U.S. infrastructure included a detailed review of capabilities in many different processes including forging, tube drawing, machining, fabrication, and testing. In addition, MPR researched the advanced manufacturing processes for additive manufacturing (i.e., 3D printing) and powder metallurgy - hot isostatic pressing. Reactor plant designers and candidate suppliers are planning to rely on traditional processes for the initial plants. If advanced manufacturing processes are developed and accepted by the nuclear industry, these processes may be used in the future. In the immediate term, ongoing research (much of which is funded by DOE) is necessary to develop and refine advanced manufacturing methods. This report includes several recommendations for DOE that will advance deployment of SMRs and advanced reactors from a manufacturing perspective and other areas identified during the course of this project.

36 MATERIALS SCIENCE↗