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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The performance of additively manufactured Haynes 282 in supercritical CO 2

The use of supercritical carbon dioxide (sCO 2 ) as a working fluid is garnering interest in next generation power production systems due to the possibility of increased operational efficiencies and lower associated costs. Its implementation requires alloys with excellent high temperature strength and corrosion resistance, which makes Haynes 282® (H282) a suitable candidate. With increasing adoption of additive manufacturing (AM) in the energy industry, there is a need to investigate long-term sCO2 exposure on AM produced materials. Additively manufactured H282 (AM-H282) samples built using laser powder bed fusion in two different orientations were analyzed through imaging, gravimetric analysis, and room-temperature tensile testing before and after 1000-h exposure to CO 2 at 750 °C and 20 MPa. Performed imaging included observation of the sample surfaces and of the material bulk (cross sections) through means of Scanning Electron Microscopy (SEM) and optical microscopy. In comparison to wrought H282, it was found that the exposed material exhibited a similar Cr 2 O 3 oxide protective layer about 2 μm in thickness with additional top-surface TiO 2 oxide and carbon-rich precipitates. It was also similarly observed that internal oxidation was present but limited to a depth of 20 μm, and a 1–3 μm γ’ denuded region appeared to surround all internal and surface oxidation. Thermal aging effects were noted with the precipitation of needle-like structures in the γ/γ’ matrix and a coarsening of the γ’ precipitates from 28 nm to 73 nm. Mass measurements analyzing oxide precipitation revealed a larger increase compared to wrought, representing an approximate 20% difference. Tensile testing results showed similar behavior to wrought with a slight increase in yield strength and a large reduction in elongation in the exposed samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Defects and Mechanical Behavior of Plutonium [Slides]

Defects are important because they heavily influence material properties (e.g. strength, ductility, elasticity, conductivity, phase transformation temperatures (melting point), corrosion resistance, etc.). At LANL defects in plutonium are being characterized using many different methods (e.g. EXAFS, XRD, microscopy, RUS, DSC, dilatometry, density, etc.) My goal is to introduce a complimentary method of understanding defects in δ-Pu using the mechanical microscope concept. Using well-developed theories of first principles strengthening mechanisms (i.e. how defects influence strength) and carefully designed experiments, mechanical testing can compliment other defect characterization tools.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Electrochemical Evolution of Fuel Cell Platinum Nanocatalysts on Carbon Nanotubes at the Atomic Scale

The evolution of Pt nanoparticles supported on carbon nanotubes is analyzed before and after electrochemical potential cycling, using identical location aberration-corrected transmission electron microscopy, for applications in proton exchange membrane fuel cells. The work is focused on the half-cell accelerated stress test protocol of potential cycles ranging between 1.0 and 1.5 V RHE to represent the start-up/shutdown settings of a fuel cell vehicle. The research work reveals that particle migration and coalescence are key mechanisms for a reduction in the Pt nanoparticle surface area at the early stages of potential cycling. Further, the mechanism for particle movement and coalescence is attributed to carbon corrosion, catalyzed either by Pt or by bulk corrosion of the carbon nanotubes. Carbon corrosion results in the appearance of carbon vacancies at the carbon nanotube/Pt nanoparticle interface during cycling, as well as the formation of edge and surface defects. During cycling, the concentration of the dissoluble Pt increases. As soon as a significant amount is reached, subnanometer/atomic clusters emerge on the carbon nanotube support, which can move and coalesce, or redeposit on the surface of larger particles through Ostwald ripening.

25 ENERGY STORAGE↗
DOE OSTI2026

Flaw Tolerance Assessment for DOE Standard SNF Dry Storage Canisters - 26550

The U.S. DOE has designed four spent nuclear fuel (SNF) dry storage canisters for storing DOE standardized SNFs. The DOE standard canisters are cylindrical shells with a diameter of 24 inches (610 m) or 18 inches (457 m), a wall thickness of 0.5 inches (12.7 m) or 0.375 inches (9.53 m), and a length of 15 feet (4.57 m) or 10 feet (3.05 m). These DOE canister geometries are completely different from commercial canisters. The latter may experience chloride-induced stress cracking corrosion (CI-SCC) because they are stored near coastal regions. The former may not experience CI-SCC but face different challenges because they are stored in the SNF storage facilities. Because of large residual stresses, mechanical flaws may occur in the DOE canisters during long-distance transportation or lifting handling. To date, only limited structural integrity analyses were carried out through drop tests on the DOE canisters, but a more general flaw tolerance assessment has not been performed. Therefore, the failure assessment diagram (FAD)-based fracture mechanics method, as codified by the latest API 579-1/ASME FFS-1-2021 Edition, is adopted in this work to assess surface flaw tolerance for DOE canisters under operation loading and welding residual stresses (WRS), where the new code-recommended WRS distributions are used. To more adequately consider the transverse distribution of WRS, an equivalent WRS distribution is proposed to account for the WRS reduction with distance from the weld centerline. Moreover, the closed-form solutions of stress intensity factor K, which serves as the crack driving force during subcritical crack growth, are developed from the tabular data of the K factors provided in API 579-1/ASME FFS-1 and used to determine more accurate flaw sizes at flaw instability. Subsequently, the Level 2 assessment procedures with 12 assessment steps, as codified and detailed in API 579-1 and ASME FFS-1, are followed to assess the flaw tolerance for the surface flaws in the DOE standard canisters with consideration of normal or accident operation loads combined with WRS. The assessment results show that the four designs of DOE standard canisters can tolerate all surface flaws that meet the code permitted maximum sizes of a flaw length of 8 inches (i.e., 200 mm) and a flaw depth of 80% wall thickness. This demonstrates that all designs of DOE standard canisters are robust and reliable.

DOE standard canister↗
DOE OSTI2020

Drying Wet Stored and Corroded Magnox Fuel for Interim Dry Storage - 20187

To investigate the feasibility of drying corroded Magnesium alloy clad nuclear fuel, the corrosion products of inactive and unirradiated Magnox simulant dry corrosion product and dried corroded Magnox sludge were characterised by X-ray diffraction and thermogravimetric analysis. XRD identified brucite in the dried corrosion product and both brucite and hydromagnesite in the sludge. TGA identified mass loss on dehydration of ∼9.1% up to 250 deg. C and 27.9% from dehydroxylation up to 440 deg. C, with total mass loss of 41.9% up to 800 deg. C for the dried corrosion product. The sludge TGA showed 8.3% mass loss up to 250 deg. C from dehydration, 7.5% up to 330 deg. C from dehydroxylation and 31.4% mass loss up to 47.2% for decarbonation, with total mass loss of 49.2% up to 800 deg. C. Water removal up to 1.8 g was performed by cold vacuum drying (40 deg. C-120 deg. C) on a ≅14.8 g (wet) corroded Magnox sample. The process was monitored by observing pressure, dew point, temperature and gas flow changes supported by measuring the sample mass loss as water is removed. From these tests, it was observed that the dried corroded Magnox displayed some hygroscopic capacity which has implications for water retention following vacuum drying and the length of exposure required to achieve the desired dryness. Whilst at higher temperatures the water was removed faster, the increased temperature also risks the fuel cladding undergoing unwanted chemical reactions with the residual water during the process. However, at lower temperatures the achievable levels of dryness were reduced. Therefore, from these observations this work has identified that there is a temperature balance that may be necessary to optimise the drying process with respect to allowing the greatest level of dryness whilst restricting unwanted chemical reactions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Laser glazing of cold sprayed coatings for the mitigation of stress corrosion cracking in light water reactor (LWR) applications

The effect of laser power, traverse speed, and cold spray coating thickness were examined with the goal of mitigating and repairing stress corrosion cracking (SCC) in light water reactor (LWR) environment. For this purpose, SCC-susceptible Alloy 600 substrate material was coated with SCC-resistant Alloy 690 via the cold spray technique. The cold spray coated substrate was then laser-glazed using various laser parameters. Single pass and multiple pass laser glazed regions were created to determine the effects. For this purpose, the area of the fusion zone, depth of the fusion zone, and elemental composition of the cross section were examined by optical microscopy, scanning electron microscopy, and energy dispersive spectroscopy. Additionally, it was shown that chromium dilution from the laser glazed region into the substrate is not a significant factor in laser glazed cold spray coated samples. Separate laser glazed samples were subjected to ASTM 633C to determine coating adhesion. Finally, by using interrupted SCC crack growth rate tests, it was shown that low traverse speeds and high laser powers produce deep fusion zones within the cold sprayed coating and substrate that can be used to seal pre-existing cracks, thus stopping SCC growth.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Performance of hydrophobic physical solvents for pre-combustion CO 2 capture at a pilot scale coal gasification facility

Here, in this paper, we present the first pilot plant data for hydrophobic physical solvents for CO 2 and H 2 S removal from coal-derived H 2 -rich syngas. Four physical solvents were tested under pre-combustion CO 2 capture conditions at bench scale and pilot plant scale: one baseline hydrophilic solvent and three hydrophobic solvents. The solvents were: (1) polyethylene-glycol-dimethyl ether (PEGDME), a hydrophilic solvent analog for the commercial process Selexol, (2) tributyl- phosphate (TBP), a commercially available hydrophobic solvent, (3) polyethylene glycol-poly(dimethylsiloxane) (PEG-PDMS-3), and (4) diethyl sebacate (CASSH-1), a novel, computationally screened hydrophobic solvent developed by the National Energy Technology Laboratory (NETL). All solvents were studied under pure gas (CO 2 /N 2 /H 2 /CH 4 ) equilibrium conditions at NETL followed by pilot plant testing with syngas at the University of North Dakota Energy & Environmental Research Center (UND EERC). Long term performance of CASSH-1 and PEDGME was then assessed with results compared to process simulation predictions. Within experimental uncertainties, all solvents showed comparable CO 2 absorption performance at above room temperature operation while the hydrophobic solvents had limited water uptake and low vapor pressure, which alleviates concerns related to corrosion, water absorption, and solvent loss to evaporation. These results indicate low viscosity, low vapor pressure hydrophobic solvents are a promising option for lower cost CO 2 capture from high pressure syngas applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Design and performance of a variable gap system for thermal conductivity measurements of high temperature, corrosive, and reactive fluids

We report high-temperature fluids such as molten salts, liquid metals, and gasses are being proposed for many advanced energy systems including thermal energy storage devices, concentrating solar plants, and advanced nuclear reactor designs. However, the chemical behavior and thermophysical properties of many of these fluids have not been well characterized, which hinders the design, modeling, safety analysis, and deployment of these systems. Thermal conductivity is a property that is especially limited by existing measurement capabilities, which are subject to errors caused by convection, material interaction, radiative heat transfer, and instrument degradation. Therefore, there is a lack of standard, systematic measurement techniques for high-temperature, reactive, and corrosive fluids. In this work, the development of a variable gap thermal conductivity measurement system is detailed. The system is designed to measure the thermal conductivity of highly corrosive and reactive fluids, and survive operation between 100 °C and 800 °C. The effects of convection are minimized by limiting the thickness of the specimen to thin sizes (<0.3 mm). Corrections for radiative heat transfer were included in the working equations to consider specimens with varying optical properties. The design, construction, instrumentation, operating principles, and data analysis techniques are discussed in detail. The system was tested up to 500 °C using helium gas and molten KNO 3 -NaNO 3 to verify the measurement technique and determine the sources of error. At 300 and 400 °C KNO 3 -NaNO 3 , results showed maximum relative error of 6% when compared to results in the literature. The helium results were within 13% of those in the literature at 300 and 400 °C. Higher errors were observed at 500 °C for both fluids, and the sources of these errors are discussed.

42 ENGINEERING↗
DOE OSTI2020

Fracture toughness evaluation for thin-shell stainless steel weldment

Integrity of spent nuclear fuel (SNF) interim storage canisters are very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method of small specimen approach was used to estimate the canister weldment fracture toughness. Most SNTT weldment samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness $J_Q'$ for the baseline SS304 steel is at 283 kJ/m 2 . The estimated $J_Q'$ for the SS304/308 weld from HAZ and weld specimens are 126 kJ/m 2 and 148.9 kJ/m 2 , respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Fracture toughness evaluations for spent nuclear fuel dry storage canister welds and spent nuclear fuel clad-pellet structures

Integrity of spent nuclear fuel (SNF) interim storage canisters is very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method was used to estimate the canister weldment fracture toughness. SNTT was developed to measure the intrinsic fracture toughness of structural materials using small specimens. The SNTT method has been applied to a wide variety of structural materials, such as low-alloy steels, stainless steel, aluminum alloy, ceramics, concrete, and composites. The SNTT system operates by applying pure torsion to cylindrical specimens with a notch line that spirals around the specimen at a 45° pitch. In order to carry out pure torsion load mode, biaxial tension/torsion tester was developed accordingly to perform SNTT protocol. Moreover, applying fracture mechanics approach to SNF system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Furthermore, due to thin wall and small dimension of clad tubing structure, the SNTT method was used to estimate the clad tubing structure fracture toughness. Fracture testing were performed on the received stainless steel canister weldment, most SNTT weld samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness J Q ’ for the baseline SS304 steel is at 283 kJ/m². The estimated JQ’ for the SS304/308 weld and baseline metals are 148 kJ/m 2 and 459 kJ/m 2 , respectively. Out of cell fracture testing for spent fuel structure were carried out on the surrogate rods made of Zr-4 clad and alumina inserts, the estimated fracture toughness values for baseline Zr-4 cladding with alumina-pellet inserts are: (1) For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate J IQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m 2 with 2-sigma uncertainty of 3.26 kJ/m 2 , and the associated K IQ is at 67.46 MPa$\sqrt{m}$; and (2) For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I - tensile stress and Mode III - out of plane shear stress; the estimated J MQ is at 18.9 kJ/m 2 , the associated K MQ is at 41.4 MPa$\sqrt{m}$.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Determination of Water Content in Bio-Oils by Volumetric Karl Fischer Titration: Laboratory Analytical Procedure (LAP)

The water content of bio-oils is a key metric for several reasons. Water is typically the most concentrated single component of fast pyrolysis oils, though this will be much lower in catalytic fast pyrolysis and upgraded products. Reduced water content is preferable in bio-oil as water can contribute to phase separation, corrosivity, and instability. Additionally, lower water is beneficial for physical properties such as energy density. The percentage of water can also be used to correct the calculation for organic oxygen when determining total oxygen content of bio-oils via combustion-based ultimate analysis. The procedure described here is specifically written for the analysis of bio-oils, and is based on ASTM E203, Standard Test Method for Water Using Volumetric Karl Fischer Titration. This standard test method is specified for measuring the mass % water in pyrolysis liquid biofuels in ASTM D7544, Standard Specification for Pyrolysis Liquid Biofuel. Although the standard method is prescribed for the analysis of pyrolysis liquids for use in industrial and commercial burners, the method is not specifically written for this product. The method described with this Laboratory Analytical Procedure (LAP) provides specific guidance for the analysis of bio-oils. Both organic and aqueous phases can be measured with this technique.

09 BIOMASS FUELS↗
DOE OSTI2023

Downselection and Basic Properties of Additively Manufactured ODS Alloys

In the AMMT program the work package for development of new materials aimed to explore the advanced manufacturing techniques that are feasible for the manufacturing of advanced materials components. A merit-based feasibility study was attempted to identify an accelerated development path for dispersion strengthened (ODS) alloys by creatively combining additive manufacturing (AM) technologies with the recent advances in ODS materials and traditional manufacturing technologies. For the fiscal year, the primary tasks were focused to develop additive manufacturing (AM) and post-build processing route for oxide dispersion strengthened (ODS) ferritic (Fe-Cr alloy or 14YWT) and austenitic (Fe-Cr-Ni alloys or 316L and 316H) alloys and to perform basic microstructural and mechanical characterizations to provide feedbacks to the alloy and processing design. Further, the multi-laboratory effort created collaboratively a decision criteria matrix to evaluate and downselect the new materials processed by advanced manufacturing methods. This report is to describe the ODS alloy processing routes combining AM processes and post-build thermomechanical treatments, mechanical and microstructural characteristics of the newly developed materials, and the application results of the decision criteria matrix for the AM ODS alloys including a downselected material and feasible processing route. Key mechanical test results including tensile strength, tensile ductility, and fracture toughness data and are reported and explained. Higher strength was measured from the ferritic ODS alloys, while higher ductility and fracture toughness were measured from the austenitic ODS alloys. Many of the decision criteria were scored the same for the AM ferritic and austenitic ODS alloys, however, the generally known fact that the austenitic alloys have higher corrosion resistance and significantly better ductility. Although these scores are not significantly different to make them highly discernable, the austenitic ODS alloys were downselected to be the primary materials group in the future research on ODS materials in AMMT program.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Drying of Spent Nuclear Fuel: Considerations and Examples

A recently revised American Society for Testing and Materials consensus standard guide for drying of spent nuclear fuel (SNF) provides considerations and guidance for preparing SNF for its safe storage in a sealed dry storage system. The standard discusses (1) needs for drying, (2) techniques to dry, and (3) demonstration of adequate dryness. No specific approach is prescribed since the effective techniques and needs for drying depend on the specifics of the fuel and dry storage system. Here, this paper discusses these topics using examples for both Zr-alloy-clad commercial SNF and for aluminum-alloy-clad research reactor SNF. Residual water can include free water (liquid and/or vapor), physisorbed water bound to internal surfaces, and chemisorbed water incorporated into surface films, such as (oxy)hydroxides. The potential impacts of these residual waters are corrosion/oxidation, radiolytic breakdown into gaseous and/or reactive species, and canister pressurization. For commercial SNF, inadvertent free water, even up to large amounts (e.g., 10+ mol), is not predicted to cause adverse corrosion degradation, except possible fuel oxidation for breached SNF. For aluminum-clad SNF, the production of radiolytic hydrogen with contribution from the chemisorbed water in its hydrated oxides is a primary consideration. For both SNF systems, canister pressurization is predicted to be well within the canister design, and flammability would not pose a safety concern using an oxygen limit of 5 vol % criterion. (Flammability control can be achieved by limiting either hydrogen or oxygen, and an oxygen limit is expected to be easier to meet in the presence of radiolytic H 2 generation.) The two primary technologies for SNF drying, vacuum drying and forced-gas dehydration, are described herein, and drying tests and campaigns using these methods are cited. Dryness criteria and the methods used to detect and measure residual (free) water are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Comparative Critical Mass Calculations for NNL and ENDF/B-VIII.0 Zirconium Hydride Thermal Scattering Laws [Slides]

Zirconium Hydride (ZrH x ) is a moderator material for TRIGA reactors and SNAP-10A.The thermal scattering has the behavior of quantum oscillator favorable for moderator feedback. ZrH x occurs as a corrosion product in LWR. Previously new phase specific TSL evaluations were generated and submitted to ENDF/B- VIII.1. These include the effects of crystal structure on elastic scattering of Zr(ZrH x ), permit flexibility in stoichiometry for use in reactor physics calculations and fuel characterization, and expand the TSL material sub-library. ENDF/B-VIII.0 ZrH TSLs do not distinguish between material phase. Current work involves preliminary validation & verification testing on the effect of TSLs on critical mass and associated thermal flux.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part I—Analysis Methods

Use of corrosion-resistant cladding can greatly extend the design life of structural components in many advanced reactor systems. However, there are currently no ASME design rules for cladded components to guard against creep-fatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This paper, presented in two parts, addresses this gap by proposing a design strategy for cladded components that do not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials—a soft clad that creeps much faster than and has lower yield stress than the class A base material and a hard clad that creeps much slower than and has higher yield stress than the class A base material. Part I discusses the approximate analysis strategies for the clad materials—treat a soft clad as perfectly compliant and a hard clad as linear elastic—and Part II develops a complete set of design rules for each of the two types of cladded components. In conclusion, finite element analyses of representative high temperature reactor components show that the proposed design analysis methods can bound the design quantities in soft cladded components and approximate the design quantities in hard cladded components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Assessment of Carbonated Brine Injection as Low-risk Strategy for Geologic Carbon Storage

The main objective of this early-phase research was to evaluate the techno-economic feasibility and risk associated with combined brine and CO 2 storage in SWD wells using brine dissolution in the North Dakota portion of the Williston Basin. Three simulation studies were conducted to investigate: (1) CO 2 phases at different conditions, (2) wellbore compatibility, and (3) long-term storage fate in reservoir. (1) A simple reservoir model and injection simulations were created using data to represent the BEST (brine extraction and storage test) site, an operational SWD facility located near Watford City, North Dakota. The pressure evolution caused by CO 2 comingled in produced water injectate in a layer cake reservoir was then modeled while tracking aqueous CO 2 throughout the project. The salinity of the injection water, the salinity of the reservoir brine, and the amount of dissolved CO 2 comingled in the injection water were varied. (2) A wellbore corrosion model was performed using the CO 2 concentrations selected based on the reservoir modeling to examine the carbonated produced water impact on wellbore. (3) Reactive transport modeling was conducted with the optimal CO 2 concentration for this injection site to study the rock-fluid interactions and CO 2 fate in the reservoir. Results suggest that CO 2 dissolved in produced water can be injected without appreciably increasing subsurface pressure or leakage risks. Pressure buildup was found to vary with salinity but not with CO 2 mass fraction. Simulation results show that lower CO 2 percent mass fraction leads to a higher amount of CO 2 that can be dissolved at a higher injection salinity. Furthermore, the long-term goal of dissolution trapping in a traditional carbon storage project is accomplished from the start, mitigating risks associated with potential migration of buoyant CO 2 , so long as the reservoir pressure and temperature are used to determine the maximum mass fraction of the dissolved CO 2 .

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Gen3 CSP Materials: Critical Review of Limited Existing and New Survey Data

Abstract Novel high temperature (≥ 700°C) Heat Transfer Medias (HTMs, e.g., molten salts) and corrosion-resistant Containment Materials (CMs, e.g., metal alloys or ceramics) are necessary for concentrated solar power (CSP) given the emphasis on higher temperatures and high cycle efficiency in the 3rd generation CSP (Gen3 CSP) technologies. In early 2019, we sent out an online survey to the Gen3 CSP community to fully assess the communal needs for thermophysical properties measurements of which HTMs and CMs, and what temperature range and other testing environments would be ideal for those materials. Based on the recorded responses, seven unique HTMs and twenty-six unique CMs were identified. Since then the list has been constantly updated, following our interactions and inputs from the Gen3 CSP community, with some new materials substituting their older counterparts. Currently, there are total of ten unique HTMs and twenty-nine unique CMs that are under consideration by the Gen3 CSP community. By analyzing the available body of research to date and combining it with our survey data from within the Gen3 CSP community, this paper presents trends of what people in the CSP world are thinking regarding materials worth investigating and suggests which thermophysical property measurements are critical to advance high-temperature CSP systems.

Gunawan, Andrey↗
DOE OSTI2021

Oxidation Behavior Comparison of UAM and Conventional Zry-4

Initial hydrothermal corrosion and steam exposures have been completed on ultrasonic additive manufacturing (UAM) Zry4 specimens to compare to wrought Zry4 performance in both normal operation and accident conditions. Slightly higher mass gains were found for the UAM Zry4 specimens in both conditions, but these were generally associated with internal oxidation where delamination of build layers occurred in the UAM material. Further characterization is being completed to conclude this first phase of testing while a second phase is being planned.

36 MATERIALS SCIENCE↗