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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Thermal cycling-driven microstructural changes of eutectic Al–Si phase change materials in SS304 containers revealed by multi-modal imaging

Aluminum-based Al–Si alloys are widely used as phase change materials (PCMs) in thermal energy storage (TES) systems owing to their high volumetric latent heat and superior thermal conductivity. However, their long-term reliability is limited by degradation processes that remain insufficiently understood. In this work, we employ a multimodal, correlative characterization framework to systematically resolve the degradation behavior of eutectic Al–Si PCMs in contact with SS304 containers under repeated thermal cycling. By integrating high-resolution electron microscopy, three-dimensional X-ray fluorescence (3D XRF) imaging, and differential scanning calorimetry (DSC), we directly link spatially resolved compositional and microstructural evolution to changes in thermophysical properties. The correlative analysis reveals that elemental leaching of Fe, Cr, and Ni from the stainless-steel container into the PCM drives the formation of intermetallic compounds (IMCs) both at the interface and within the bulk PCM, leading to pronounced compositional heterogeneity. These interfacial reactions and diffusion-induced transformations progressively destabilize the Al–Si eutectic, reducing the effective phase-transforming fraction. Consistent with these observations, DSC measurements show a decrease in melting temperature and latent heat of fusion with thermal cycling. These results underscore the critical influence of interfacial reactions and material compatibility on the stability, durability, and overall performance of Al–Si-based TES systems.

25 ENERGY STORAGE↗
DOE OSTI2013

Advanced Instrumentation for Extreme Environments

The Department of Energy (DOE) Office of Nuclear Energy (NE) is pursuing embedded instrumentation and controls (I&C) technology for next generation nuclear power generation applications. Embedded systems encompass a wide range of configurations and technologies; we define embedding in this instance as the integration of the sensors and the control system design into the component design using a systems engineering process. Embedded I&C systems are often an essential part of developing new capabilities, improving reliability, enhancing performance, and reducing operational costs. The new intrinsically safe, more efficient, and cost effective reactor technologies (Next Generation Nuclear Plant and Small Modular Reactors) require the development and application of new I&C technologies. These new designs raise extreme environmental challenges such as high temperatures (over 700 C) and material compatibility (e.g., molten salts). The desired reliability and functionality requires measurements in these extreme conditions including high radiation environments which were not previously monitored in real time. The DOE/NE Nuclear Energy Enabling Technologies (NEET) program currently has several projects investigating I&C technologies necessary to make these reactor designs realizable. The project described in this paper has the specific goal of investigating embedded I&C with the following objectives: 1.Explore and quantify the potential gains from embedded I&C improved reliability, increased performance, and reduced cost 2.Identify practical control, sensing, and measurement techniques for the extreme environments found in high-temperature reactors 3.Design and fabricate a functional prototype high-temperature cooling pump for molten salts represents target demonstration of improved performance, reliability, and widespread usage There are many engineering challenges in the design of a high-temperature liquid salt cooling pump. The pump and motor are in direct contact with molten fluoride salt at 700 C (1,292 F) as part of a reactor cooling loop. The motor-pump combination during normal operation would be red-hot (Figure 1). This environment challenges every facet of the design including seals, wiring, magnetic materials, and sensors. In this paper, we discuss the challenges of sensor design in extreme environments and specifically the sensor design for a high-temperature fluoride salt coolant pump. This pump will be used as a test-bed for embedded I&C development and validation in extreme environments.

Extreme Environments↗
DOE OSTI2021

Embedded Sensors in Additively Manufactured Silicon Carbide

Silicon carbide (SiC) components are being considered for a wide range of nuclear applications due to their high-temperature strength retention, low neutron absorption, chemical inertness, and dimensional stability under neutron irradiation. However, machining and joining of SiC components have traditionally limited its application to relatively simple geometries. Recent work has demonstrated additive manufacturing of complex, high-purity, crystalline SiC components using a combination of binder jet printing and densification via chemical vapor infiltration (CVI). The process lends itself to embedding of fuel, absorbers, moderators, and sensors at strategic locations within a component. The latter could allow for enhanced in situ performance monitoring of limiting fuel temperatures, self-shielded neutron flux, and potentially spatially distributed strain within complex SiC components if sensors can be successfully embedded during CVI. This work describes (1) methods for embedding sensors; (2) thermodynamic analyses and material compatibility testing for identifying sensors capable of surviving high temperatures and exposure to hydrogen and hydrogen chloride during CVI; and (3) nuclear applications for embedded sensors, including potential failure modes during fabrication and during reactor operation. Molybdenum-sheathed thermocouples were successfully embedded in a complex SiC component, whereas niobium-sheathed high-temperature irradiation-resistant thermocouples started to drift as soon as the reactant gases were introduced and ultimately failed during CVI due to severe constrained expansion, potentially resulting from niobium hydride formation in the low-temperature region of the CVI system. Optical fibers were successfully embedded in SiC, but further work is needed to protect the fragile fiber leads after their protective coatings are removed during CVI.

thermocouples↗
DOE OSTI2025

Roadmap for Assessing Fuel Reprocessing in Fluoride-Based Salts

This joint report assesses the pyroprocessing of used nuclear fuel from molten fluoride salts being conducted between Idaho National Laboratory and Argonne National Laboratory. The goal of this report is to identify the research and development gaps needed to reduce the technical risks of extending pyroprocessing technologies and unit operations to molten fluoride salts used in molten salt reactors (MSRs). Assessing and performing the tasks outlined in this report will help mitigate the technical risks and design appropriate flowsheets for reprocessing fuel and coolant salts. These suggested tasks will provide necessary data to implement the development of unit operations. Sections are highlighted to identify processes that need to be assessed and unit operations that may be used in fluoride-based chemistries. Research and development needs are listed and discussed including re-fluorination methods, electrowinning for separations, reference electrode development, materials compatibility, salt purification/re-fluorination, and waste disposition. This report will examine the technical challenges associated with pyroprocessing in molten fluoride-based salt and outline approaches to increase the technical readiness level of pyroprocessing in molten fluoride salts.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mitigating Electronic Conduction in Ceria‐Based Electrolytes via External Structure Design

Doped ceria electrolytes are the state of the art low‐temperature solid oxide electrolytes because of their high ionic conductivity and good material compatibility. However, cerium tends to reduce once exposed to reducing environments, leading to an increase in electronic conduction and a decrease in efficiency. Here, the leakage current is mitigated in ceria‐based electrolytes by controlling the defect chemistry through an engineered cathode side microstructure. This functional layer effectively addresses the problematic electronic conduction issue in ceria‐based electrolytes without adding significant ohmic resistance and increases the ionic transference number to over 0.93 in a thin 20 µm ceria‐based electrolyte at 500 °C, compared to a of 0.8 for an unmodified one. Based on this design, solid oxide fuel cells (SOFCs) are further demonstrated with the remarkable peak power density of 550 mW at 500 °C and excellent stability for over 2000 h. This approach enables a potential breakthrough in the development of ceria‐based low‐temperature solid oxide electrolytes.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Investigation Into LiCl-LiF+Li 2 O as a Low Temperature Electrolyte in Direct Electrolytic Reduction of UO 2

Direct electrolytic reduction (DER) of UO 2 in a binary LiCl-Li2O (1–3 wt%) at 650 °C has been widely reported. The process temperature can be reduced to 550 °C in theory by using a near eutectic ternary LiCl-LiF+Li 2O (∼2 wt%) electrolyte, which will result in substantial reduction in rate of salt vaporization. The feasibility of using the ternary electrolyte for the process was tested via material compatibility testing, electrochemical polarization testing, and controlled potential or current DER experiments. MgO, Pt, and 625 nickel samples were each contacted with the ternary salt for 168 h. Negligible changes to dimensions and mass of the samples were measured, and concentrations of corrosion products in the salt were 102 ppm or less. O2−oxidation was observed to occur at a lower potential than Cl-or F-on a Pt anode. DER of UO2was performed in six trials. Process variables included type of reference electrode, electrochemical control method, type of cathode basket, and electrolyte composition. UO 2conversion ranged from 28.2 to 99.9 % , and current efficiency ranged from 38 to 52%. Overall, results indicate that the use of the ternary salt is feasible for DER.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Material properties of aqueous potassium acetate solutions for use in liquid desiccant air conditioning

We present thermophysical properties of aqueous potassium acetate solutions relevant to their application as a liquid desiccant in air-conditioning systems. Liquid desiccant air conditioners could provide up to 80% energy savings compared to high-efficiency vapor compression air conditioners. However, the commonly used liquid desiccants, particularly chloride salt solutions, are highly corrosive. This precludes the use of metallic components, necessitating specialized plastics and thereby driving up cost, weight, and limiting operational temperature and pressure ranges. Potassium acetate has been identified as a less corrosive alternative, potentially enabling broader material compatibility in LDAC components. In this study we systematically measured and modelled key thermophysical properties of KAc solutions density, vapor pressure, viscosity, specific heat capacity, phase behavior, and diffusion coefficient across a range of concentrations (10 - 70wt%) and temperatures. For instance, KAc exhibits a vapor pressure of 3.1 kPa at 50wt% and 40degrees C, demonstrating dehumidification capability comparable to chloride salts. By providing an extended dataset and associated models, this work advances the understanding needed to design energy-efficient, corrosion mitigating LDAC system.

36 MATERIALS SCIENCE↗
DOE OSTI2013

A Review of Tungsten Heavy Alloy Utilization in Isotope Transport Containers - 13380

A common requirement for radioisotope transport containers is that they provide both durable and efficient shielding of penetrating gamma radiation. This is the case for transport of both spent nuclear fuel as well as intentionally created radioisotopes for medical or other uses. Tungsten heavy alloy (WHA) provides a unique engineering property set for such shielding - easily surpassing more commonly used lead alloys in both strength and attenuation. This family of alloys contains typically 90-98 wt.% W in combination with transition metals such as Ni and Fe. WHA is manufactured in near net shape blanks by liquid phase sintering of compacted powder shapes to full metallurgical density parts. This powder metallurgy approach is described in its ability to provide excellent material utilization and affords efficient manufacturing of various shapes required for gamma shields or collimators. WHAs offer very high density (approaching 19 g/cc) in combination with relatively high thermal conductivity, low thermal expansion, ambient corrosion resistance, and can be provided with mechanical properties comparable to many medium carbon steels. As such, they can be machined to complex, damage resistant geometries using common metal cutting tools and methods. WHA additionally provides a lower toxicity alternative to Pb- or U-based gamma shielding. Given the specialty nature of WHA, specific metallurgical characteristics are reviewed to assist shielding designers who may otherwise encounter difficulties locating important alloy selection and fabrication details. Contained within this materials and applications overview are guidelines for WHA component design, alloy selection, and practical machining, finishing, and assembly considerations. The microstructure of WHA is that of a metal matrix composite. This factor has specific implications in the design of components for stress service as well as their protection in the presence of electrolytes. WHA is also discussed in the broader context of materials compatibility, as it is rarely used in isolated monolithic shapes. Alloy selection for new applications is often made primarily on the basis of density. An alternative strategy to this selection approach is presented which proposes that mechanical requirements for a given shielding use be the primary selection criterion over density. Standard commercial grades of WHA for radiation shielding are defined by specifications such as AMS-T-21014 and ASTM B777. These specifications define 4 density classes of as-sintered WHA. Compositional options, as well as post-sinter processing of WHA, are discussed for shielding components that must exhibit higher levels of ductility or very low magnetic permeability. In addition to the mechanical advantages over Pb-based shielding, the higher linear attenuation of energetic photons for various grades of WHA (as calculated by the NIST XCOM routine) are presented for selected photon energies of interest to illustrate the shield volume reduction generally possible through the use of tungsten-based shielding. While W provides inefficient attenuation of neutrons in a mixed radiation environment, its secondary role in shielding gamma radiation produced as a result of neutron capture is also described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System

To better enable informed decision making regarding the back end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed SNF. DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified HLW similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System PPT

To better enable informed decision making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed Spent Nuclear Fuel (SNF). DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed for the packaging demonstration. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified high-level radioactive waste (HLW) similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP)

This document constitutes progress made on this project by the Georgia Tech led Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP). This is the first draft of the project final report. Chapter 3 is focused on Tritium management. Tritium management in Fluoride-salt High-temperature Reactors (FHRs) presents a major operational concern and a significant challenge. Tritium is generated via multiple pathways including various interactions in fuel and structural materials, as well as coolant (primarily interaction with lithium and beryllium), leading to a complex evolution and spatial distribution of tritium generation rate. FHR generates orders of magnitude more tritium than in LWRs, and has potential for a significant release, via permeation through structural materials (due to high temperature), and through heat exchanger(s). Therefore, accurate estimate is necessary. For that purpose relevant tritium production pathways were identified. A methodology to obtain space-time dependent tritium generation rate was developed and implemented, employing SCALE and MCNP codes. Detailed models of the reference Advanced High Temperature Reactor (AHTR) design were developed. A multi-parametric metrics was introduced to capture relevant tritium production indicators, including the instantaneous tritium production rate, total tritium production (neglecting subsequent decay), and accumulated tritium production (accounting for the decay). The obtained results will inform tritium management, including the sizing of the components, and to gain insight on the impact of the core design parameters on tritium generation rates. Research effort on Liquid Salt Impurity Removal, Redox and Corrosion Control is discussed in Chapter 4. Section 4.1 is an introduction chapter. Based on the current available knowledge and operation experience on previous MSRs and other types of nuclear reactors, the report identifies the major impurities that are related to either material compatibility, salt properties or radioactive source terms, and the potential contamination sources of these impurities, their transport in the primary coolant loop and available cleanup methodologies for a particular group of impurity. In section 4.2, the available purification methods were discussed, including the liquid/liquid extraction using liquid Bi-Li which has been applied in MSBR and electrochemical purification which is developed based on Pyroprocessing. In this section, the new multi-stage and multi-salt-component was developed and discussed. Section 4.3 discussed the redox potential control for salt chemistry and corrosion mitigation, including the control theory and the soluble/soluble buffer control method. The section also developed method on how to analyze soluble/soluble redox reaction in a molten salt system and designed an electrochemical sensor for redox potential control. Corrosion models were also reported in this section. Section 4.4 focused on how to measure the fundamental properties in a molten salt. As an example, the section focused on La fluoride in the molten salt. The section developed methods on how analyze electrochemical measurement data to obtain the fundamental properties including the redox potential, diffusion coefficient and exchange current density. All these fundamental information is important for salt purification system development. Section 4.5 focused on measurements of corrosion products in the salts. Because of the different behaviors between Cr and La in the salt, new analysis method for Cr fluorides in the salt was developed. Some results for Fe fluoride in the salt and corrosion measurements were also reported. Section 4.6 demonstrated the separation of oxide from the salt. This section also measured the solubility of La 2 O 3 in the salt as function of temperature. Section 4.7 focused on the separation of La and Ce from the molten salt. These two elements were selected because La was found to be separated by using inert electrode, while for separation of Ce, reactive electrode must be selected. Section 4.8 demonstrated the separation iodine ions from the molten salt. Two kinds of working electrode materials (metal W and graphite) were investigated. Section 4.9 focused on the fundamental data of fission products and actinides in liquid Bi. The chapter reviewed the available data, then based a CALPHAD model, the available data were broadened to the concentration-dependent cases.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Mechanistic transition of stainless steel 304L corrosion in MgCl 2 phase change material under thermal cycling

Magnesium chloride (MgCl 2 )-based molten salt phase change materials (PCMs) are promising candidates for thermal energy storage (TES) owing to their advantageous thermophysical properties, however, their high corrosivity in the molten state limits material compatibility. In this study, the corrosion behavior of 304?L stainless steel (SS304L) exposed to MgCl2 PCM was investigated as a function of thermal cycling (up to 9000 cycles) relevant to PCM-based TES systems. Microstructural changes and elemental redistribution at the alloy-salt interface under thermal cycling conditions were characterized using synchrotron-based X-ray fluorescence (XRF) and scanning electron microscopy coupled with energy dispersive X-ray spectroscopy (SEM-EDS). Results show that corrosion is governed by selective Cr-dissolution and elemental diffusion near the interface and grain boundaries. A clear transition in the dominant corrosion mechanism is observed with increasing thermal cycles. Surface corrosion is the dominant mechanism up to 1000 cycles, whereas intergranular corrosion becomes the primary degradation mode beyond 2000 cycles. These mechanistic changes are interpreted in terms of evolving interfacial chemistry, microstructure and the diffusion-driven redistribution of alloying elements.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Approaches to investigate the role of chelation in the corrosivity of biomass-derived oils

The need to provide the U.S. market with a renewable liquid fuel energy source from a non-food feedstock stream has gained considerable traction due to benefits such as improved energy efficiency, reduced environmental impacts, and enhanced national security. Practical achievement of these goals via biomass and bio-waste utilization involves production of liquid intermediates containing corrosive, reactive species like carboxylic acids, ketones, aldehydes, and hydroxyaldehydes. Such mixtures challenge materials of containment, processing, and transport. It is widely recognized that the smaller organic acids, such as acetic and formic, are corrosive and can remove protective surface oxides on alloys used in bio-oil processing infrastructure, and ketones can swell sealing polymers. However, literature shows, and findings herein confirm, larger carboxylic acids and bidentate alcohols are present. This highlights the potential for synergistic, detrimental effects of constituents in bio-oil corrosion, including direct reactivity of small acids compounded with the possibility of mobilization of protective metal oxide layers via chelation by larger acids and oxygenates. The question of whether species beyond small acids can significantly contribute to corrosion requires analytical approaches previously not applied to bio-oil corrosion studies and certainly not previously applied corroboratively. As such, this work introduces a combination of optical, mass spectral, and electrochemical impedance spectroscopies with an incubation approach to study metal mobilization, to facilitate elucidating chelation's role in bio-oil corrosive pathways. To enable systematic study of these oxygenates' material compatibility individually and in combination, a model matrix of bio-oil constituents was also developed based on identification of key components of real bio-oils.

09 BIOMASS FUELS↗
DOE OSTI2016

Parameters controlling early age hydration of cement pastes containing accelerators for sprayed concrete

The objective of this work is to parametrize the early age hydration behavior of accelerated cement pastes based on the chemical properties of cement and accelerators. Eight cements, three alkali-free and one alkaline accelerators were evaluated. Isothermal calorimetry, in situ XRD and SEM imaging were performed to characterize kinetics and mechanisms of hydration and the microstructure development. The reactivity of all accelerators is directly proportional to their aluminum and sulfate concentrations and to the amount and solubility of the setting regulator contained in cement. Alite hydration is enhanced if a proper C{sub 3}A/SO{sub 3} ratio (between 0.67 and 0.90) remains after accelerator addition and if limestone filler is employed, because undersulfated C{sub 3}A reactions are avoided. Combinations of compatible materials are recommended to enhance the performance of the matrix and to prevent an undesirable hydration behavior and its consequences in mechanical strength development.

36 MATERIALS SCIENCE↗
DOE OSTI2019

A horizontal powder injector for W7-X

We report that injection of low-Z powders into fusion plasma has been used to improve wall conditions, similar to the standard boronization process using diborane. Powder injection has the advantage of being much simpler, non-toxic, and efficient. The W7-X stellarator is planning on utilizing powder injection in long pulse discharges; a proof-of-principle test for horizontal injection into the plasma was conceived, a Probe Mounted Particle Injector (PMPI). The PMPI concept is developed using a polyetheretherketone (PEEK) paddle wheel hub that is driven by a piezo motor, due to the high magnetic fields, that rotates at 100 deg/s. This small PMPI fits into an envelope of 120 mm diameter × 150 mm long, a standard size for Multi-Purpose Manipulator (MPM). The device is housed in a carbon cup mounted on a retractable probe that can be placed near the plasma edge, enabling powder injection ˜4–8 cm radially into the boundary plasma. The feed for the paddle is via piezo electric actuator that vibrates a funnel filled with powder into a trough for the paddle to push. The 8 paddle arms, 35 mm long and 10 mm wide, are made from 0.38 mm thick PEEK which drag slightly along the powder-filled trough bottom, becoming a spring-loaded paddle which accelerates the powder upon release. Finally, design challenges are the high ambient magnetic field, vacuum compatible materials, high temperature environment, limited rotary-drive options, and compact space. The design and testing of this new device will be presented.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Pool boiling heat transfer evaluation of next-generation dielectric fluid: Opteon™ 2P50

The growing use of artificial intelligence has led to heavy thermal loads and high heat dissipation rates in data centers. Conventional air-cooled technologies are not able to fulfill these requirements. To overcome these challenges, two-phase immersion cooling (2PIC) has emerged as one of the leading technologies for high power-density chips. 2PIC increases the heat dissipation rate and efficiency of the system while reducing the footprint of the cooling equipment. A fluid with adequate dielectric properties, a suitable normal boiling temperature to maintain chip temperatures, and good material compatibility, is desired for 2PIC system. In this study, the pool boiling heat transfer of a new developmental dielectric fluid, Opteon™ 2P50, was experimentally investigated. The heat transfer coefficients at various heat fluxes (20–150 kW/m 2 ) and the critical heat flux were measured using a smooth aluminum surface. Compared with HFE-7100, Opteon™ 2P50 shows higher heat transfer coefficient (up to 59% higher) and a slightly lower value of critical heat flux (around 5.9% lower). The modified Cooper correlation with the optimized leading constant resulted in reliable prediction accuracy with a 5.3% mean absolute error percentage. Overall, these results indicate that the new dielectric fluid provides similar thermal performance to some legacy fluids.

2P50↗
DOE OSTI2020

On the fatigue crack growth behavior of Ti–10V–2Fe–3Al in gaseous hydrogen

Here, the fatigue crack growth behavior of Ti–10V–2Fe–3Al in gaseous hydrogen (H 2 ) was assessed through comparative experiments conducted in laboratory air and 8.3 MPa H 2 . The measured fatigue crack growth rate (da/dN) versus applied stress intensity factor range (ΔK) relationships and observed fracture morphologies for laboratory air and H 2 were comparable up to ΔK ≈ 6.9 MPa√m, when tested at a load ratio of 0.1 and frequency of 10 Hz. At higher ΔK values, significant crack deflection and subsequent catastrophic failure occurred in the specimen tested in H 2 . This degradation was not observed in a specimen pre-exposed to 8.3 MPa H 2 for 96 h and then immediately tested in laboratory air. X-ray diffraction of the failed H 2 -tested specimen revealed that the material remnants were predominantly composed of TiH 2 , suggesting that hydride formation was the catalyst for catastrophic failure in H 2 . The mechanistic implications of these results and their impact on current material compatibility assessments for Ti alloys in hydrogen service are then discussed.

08 HYDROGEN↗