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At least 55 records · Page 3

Iodine Hall-Effect Electric Propulsion System Research, Development, and System Durability Demonstration

This paper reviews recent iodine electric propulsion research and development activities at the NASA Glenn Research Center (GRC). Activities included (i) investigation of the iodine compatibility of BaO-CaO-Al2O3 impregnated tungsten hollow cathodes based on a flight heritage design, (ii) investigation of the iodine compatibility of a handful of materials common to propulsion systems, spacecraft, and ground test facilities, (iii) development of reliable iodine feed system technologies, (iv) implementation of test facility improvements in an attempt to mitigate iodine associated negative impacts, and culminated in (v) an 1,174-hour hybrid iodine-xenon propulsion system durability demonstration (iodine fed Hall-effect thruster with xenon fed cathode). Each of the activities resulted in extensive insights that shall informfuture iodine electric propulsion developments. While reliable operation of a BaO-CaO-Al2O3 impregnated tungsten hollow cathode on iodine vapor was not achieved, long-term operation on xenon gas in proximity to an iodine fed thruster was demonstrated without any measurable degradation or cross-contamination of the cathode. Furthermore, iodine material corrosion investigations conducted at 300 degrees C over 5, 15, and 30 days showed significant deterioration of all materials evaluated, although the same materials with a silicon coating proved nearly impervious to iodine so long as the coating was not mechanically damaged. Finally, the 1,174-hour durability test demonstration showed that (i) iodine feed system technologies developed at GRC delivered well-regulated uninterrupted propellant, (ii) implementation of appropriate facility improvements and procedures can limit negative impacts of iodine on test hardware and ground support equipment, although facility challenges with iodine are extensive, and (iii) a Hall-effect thruster operates with similar performance whether employing iodine or xenon propellant over long durations. The work was motivated by strong government and commercial interest in the growing capabilities of small-spacecraft (<500 kg), in combination with interest for denser low-power, high delta-v in-space propulsion systems. This work adds to a growing body of research and development efforts aimed at addressing the many anticipated challenges of implementing iodine as an in-space propellant. This work was conducted under the Advanced In-Space Propulsion (AISP) project funded through the Game Changing Development (GCD) program within NASA's Science Technology Mission Directorate (STMD).

Benavides, Gabriel F.↗

Improved fracture toughness corrosion-resistant bearing material

A development program was performed to establish whether a corrosion-resistant bearing material, such as a 14Cr steel, could be modified to allow carburization, thereby providing the excellent fracture toughness characteristics feasible with this process. The alloy selected for investigation was AMS 5749. Several modifications were made including the addition of a small amount of nickel for austenite stabilization. While some promising results were achieved, the primary objective of an acceptable combination of case hardness and microstructure was not attained. Because the high chromium content presents a serious problem in achieving a viable carburizing cycle, a number of experimental steels having lower chromium contents (8 to 12%) were produced in laboratory quantities and evaluated. The results were basically the same as those initially obtained with the modified AMS 5749. Corrosion tests were performed on AMS 5749, AISI M50, and 52100 bearing steels as well as some of the lower chromium steels. These tests showed that a reduced chromium level (10 to 12%) provided essentially the same corrosion protection as the 14Cr steels.

Bamberger, E. N.↗

Improved ceramic heat exchange material

Improved corrosion resistant ceramic materials that are suitable for use as regenerative heat exchangers for vehicular gas turbines is reported. Two glass-ceramic materials, C-144 and C-145, have superior durability towards sulfuric acid and sodium sulfate compared to lithium aluminosilicate (LAS) Corning heat exchange material 9455. Material C-144 is a leached LAS material whose major crystalline phase is silica keatite plus mullite, and C-145 is a LAS keatite solid solution (S.S.) material. In comparison to material 9455, material C-144 is two orders of magnitude better in dimensional stability to sulfuric acid at 300 C, and one order of magnitude better in stability to sodium sulfate at 1000 C. Material C-145 is initially two times better in stability to sulfuric acid, and about one order of magnitude better in stability to sodium sulfate. Both C-144 and C-145 have less than 300 ppm delta L/L thermal expansion from ambient to 1000 C, and good dimensional stability of less than approximately 100 ppm delta L/L after exposure to 1000 C for 100 hours. The glass-ceramic fabrication process produced a hexagonal honeycomb matrix having an 85% open frontal area, 50 micrometer wall thickness, and less than 5% porosity.

Mccollister, H. L.↗

Development of spectroscopy apparatus for molten salt reactor safeguards and material characterization

A molten salt reactor (MSR) produces nuclear fission energy using mixture of fissile and non-fissile salts. Salt in the reactor functions as the nuclear material and the coolant allowing for a safer design. International nuclear safeguards is the monitoring of nuclear reactors, nuclear fuel, and irradiated material to prevent the proliferation of nuclear weapon grade material in hostile foreign nations. The mixture of salt creates a proliferation risk as it generates a conglomeration of fuel, daughter nuclei, and corrosive material. A method to monitor and measure the materials present in the reactor is essential to the scalable deployment of MSRs. The objective of this project was to design a spectroscopy apparatus for aqueous salt so it can be measured and collected consistently in the aerosol phase.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

High-temperature corrosion of a nickel-based alloy in a molten chloride environment – The effect of thermal and chemical purifications

There is currently an ever-increasing demand for higher process efficiencies in next generation (Gen3) concentrating solar power (CSP). Higher process efficiencies may be procured by increasing the operating temperature, and simultaneously, minimizing the degradation of materials used for construction of CSP plants (e.g., piping, thermal storage tanks, solar receivers and heat exchangers). Thus, understanding materials corrosion in the presence of molten salt mixtures used as thermal energy storage media and heat transfer fluids is indispensable for CSP development. The present paper provides insights into the effects of salt purification on the corrosion of a nickel-based alloy (Haynes 230) isothermally exposed to a stagnant chloride-based salt mixture at 800 °C. The MgCl 2 -based salt mixture was thermally dehydrated and chemically treated with (0.1 and 0.5 wt %) elemental magnesium. Results reveal the electrochemical nature of the corrosion process, and the formation of corrosion products such as oxides (MgO, MgCr 2 O 4 , and MgAl 2 O 4 ) and nitrides (CrN) on the alloy surface and in sub-surface regions. Magnesium additions enhanced the ability to resist corrosion by reducing the concentration of impurities (H 2 O, MgOH + , OH - species) and polarizing the alloy surface. The formation of nitrides in all cases studied indicates the impact of using nitrogen as a protective gas in the system. Results also reveal that a single step treatment of the salt using metallic Mg could be considered as a measure to control the salt's impurity level, e.g., if required for system control.

14 SOLAR ENERGY↗

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification - APPS1 Matrix Glasses

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification - HAL24 M1 Matrix Glasses

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Status Report on the Development of Cladding Inner-wall and Outer-wall Coatings

Advanced cladding is critical for nuclear reactors with enhanced performance in radiation tolerance, neutron transparency and reactor safety. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance and high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified outer surface, a robust clad base material, and a modified inner wall) promises to meet all these requirements. In this status report, the initial efforts on the development of cladding inner-wall and outer-wall coatings are summarized. The activities covered in this report are (1) High-resolution TEM studies of TiN/HT-9 interface atomic structure and chemistry after ion irradiation, (2) High-resolution TEM studies of TiN/14YWT ODS interface atomic structure after thermal cycling, (3) Inner-wall TiN coating for an experimental 14YWT ODS cladding, and (4) Corrosion test of TiN coated Ziracloy-4 bar samples.

36 - MATERIALS SCIENCE↗

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification – VSL Touchpoint Matrix Glasses (Revision 1)

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF).[1] The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Metal-halide mixtures for latent heat energy storage

Some candidates for alkali metal and alkali halide mixtures suitable for thermal energy storage at temperatures 600 C are identified. A solar thermal system application which offer advantages such as precipitation of salt crystals away from heat transfer surfaces, increased thermal conductivity of phase change materials, corrosion inhibition, and a constant monotectic temperature, independent of mixture concentrations. By using the lighters, metal rich phase as a heat transfer medium and the denser, salt rich phase as a phase change material for latent heat storage, undesirable solidification on the heat transfer surface may be prevented, is presented.

Chen, K.↗

Status Report on Ion Irradiation Study of Ceramic Coating in Suppressing Fuel/Cladding Chemical Interaction

Advanced cladding is critical for the development of advanced nuclear reactors with enhanced performance in radiation tolerance and neutron transparency. It will ensure adequate thermal conductivity and mechanical stability of cladding base material, corrosion resistance and high temperature coolant compatibility of cladding surface. Finally, it provides chemical stability in the cladding inner wall against fuel/cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a treated surface, a base material, and a treated inner-wall) promises to meet all the listed requirements. The initial work on innovative researching and developing these cladding materials seeks to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI through ion-irradiation. It consists of a study of Ce/TiN/Fe diffusion couples. This report provides an update on the status of ceramic-coating preparation and the testing of diffusion couple samples, the design of an ion-irradiation experiment, and characterization of the microstructure of a legacy-coupon sample coated with 1 µm of TiN and irradiated with 5 MeV Fe ions to 200 dpa at 500°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Summary Report on Ion Irradiation Study of Ceramic Coating on Suppressing FCCI

Advanced cladding is critical for advanced nuclear reactors with an enhanced performance in radiation tolerance and neutron transparency. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance, high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified surface, a clad base material, and a modified inner wall) promises to meet all these requirements. Initial research and development (R&D) regarding this innovative cladding seek to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI under ion irradiation to high dose. This United States (U.S.) Department of Energy (DOE)–Office of Nuclear Energy (NE) project report summarizes the results of recent ion irradiation studies of diffusion-couple samples with details of an ion irradiation experiment, the characterization of the interface microstructure of cerium (Ce)/titanium nitride (TiN)/oxide-dispersion strengthened (ODS), Ce/TiN/iron (Fe), Ce/ODS, and Ce/Fe samples irradiated with 80 million electron-volts (MeV) of xenon (Xe) ions to 100 displacements per atom (dpa) at 500?C using the Argonne Tandem Linac Accelerator System (ATLAS) facility at Argonne National Laboratory (ANL). The results of the ODS-substrate sample are encouraging and demonstrate the effectiveness of a thin ceramic coating on FCCI mitigation under irradiation. The results from an Fe-substrate sample revealed a complex microstructure, and the root cause is discussed.

36 MATERIALS SCIENCE↗

Investigating the Adhesion Force of Lunar Regolith Particulates on Air Plasma Sprayed Alumina Coatings

Ceramic materials, often used to protect components due to their high strength, and wear resistance, have added benefits of being lightweight and providing multifunctional properties. They are, therefore, significant to provide durability and support during long-duration missions to the moon’s surface for rovers, landers, robotic systems, habitats, and many other components. Materials such as aluminum oxide with high mechanical strength and hardness can help to improve the durability of structures used in space exploration. Devices used for space exploration require the use of materials with the ability to withstand exposure to extreme environments. Lunar regolith is a constituent with the ability to adhere electrostatically and damage the components used to perform these missions due to adhesion of lunar dust projectiles that can cause delamination on surfaces, which may not be physically visible. The particulates of lunar dust are classified as corrosive material, leading to the degradation of structures. Aluminum oxide presents excellent resistance to different types of wear due to its high strength and hardness. Air plasma sprayed (APS) aluminum oxide coatings have demonstrated the potential to protect the surface to which they are applied. However, the parameters of roughness and porosity of the coatings need to be considered to establish if they can protect the components from extreme environments. Considering the electrostatic forces that the components are exposed to, the adhesion between the lunar regolith and the surface of the components needs to be investigated. In this work, the adhesion forces of lunar dust simulants with an average size of 30 𝜇m are investigated considering the roughness of the surface of the APS aluminum oxide. The centrifugal technique utilized here offers the advantage of establishing the adhesion force between particles with different shapes on smooth or rough surfaces. The simulants were deposited on the surface of APS aluminum oxide using an aerosolization technique to achieve a monolayer coating in three different locations across the specimen to determine the increments in adhesion force at the centrifuge. The roughness of the APS aluminum oxide was determined to be 2.257 𝜇m. The specimen was tested under an incremental centrifugal speed from 100 rpm to 3000 rpm for 20 seconds. Low magnification microscopy images were collected to cover a larger surface area of the test coating. The adhesion force was measured considering the distance from the centrifugal axis and the rotational speed. Due to the coating roughness, and the distance from the centrifugal axis, the results demonstrated a larger adhesion force in locations closer to the centrifugal axis. Smaller particles were entrapped within the roughness of the coating, and a stronger adhesion force was measured. In order to continue designing wear-resistant coating for structural protection in space missions, press-on experiments using a centrifuge will be performed. Future experiments will allow us to determine the design parameters for APS aluminum oxide coating to protect the structures from harsh space environments.

Perla Latorre-Suarez↗

High-temperature chromium diffusion in austenitic stainless steel: Ab initio molecular dynamics simulations

Chromium self-diffusion through stainless steel (SS) matrix and along grain boundaries is an important mechanism controlling SS structural materials corrosion. For this work, Cr diffusion in austenitic SS was simulated using canonical ab initio molecular dynamics with realistic models of type-316 SS bulk, with and without Cr vacancies, and a low-energy Σ3 twin boundary typically observed at active corrosion sites. Cr self-diffusion coefficients at 750 and 850 °C calculated using Einstein’s diffusion equation are 4.2 × 10 –6 and 8.1 × 10 –6 Å 2 ps –1 in pristine bulk, 3.8 × 10 –3 and 5.5 × 10 –3 Å 2 ps –1 in bulk including Cr vacancies, and 9.5 × 10 –2 and 1.0 × 10 –1 Å 2 ps –1 at a Σ3[1 1 1]60° twin boundary.

36 MATERIALS SCIENCE↗

Description of FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

The urgency to deliver fusion power is growing now more than ever, with increasing pressure for both public programs and private companies to meet milestones timelines and overcome significant remaining technical challenges to ensure growth of a nascent fusion industry in time to meet rapidly growing clean energy demands. With incredible advancements in computation and years of investment in fusion model development and validation, integrated modeling is poised to fill a key role in accelerating the timeline to a fusion pilot plant (FPP). Future fusion pilot plants will operate in regimes far beyond current experience, and device design will rely on physics-based prediction and extrapolation. Many concepts will also rely on simulation to assess safety (shielding, tritium management, materials activation and lifetimes), economics and scalability before the decision to build. Importantly, integrated simulation can be used to reveal and solve the complexities of system integration that may otherwise not be apparent in physical components or models developed in isolation. New experimental test facilities that produce relevant conditions to validate and resolve key technical challenges for various subsystems (materials, blankets, fuel cycle, etc.) have been repeatedly called for by the fusion community but are not yet realized. Integrated modeling has an important role in identifying realistic load conditions (thermal, electromagnetic, plasma, neutron and photon loads, etc.) and defining the components and experiments for these test facilities in order to ensure meaningful validation that sufficiently reduces modeling uncertainties and technical risk for the full integrated reactor. The Fusion REactor Design and Assessment (FREDA) SciDAC project is building a component-based integrated modeling framework & data structure to enable self-consistent, multi-fidelity, iterative optimization workflows for the fusion reactor design process. FREDA aims to shorten the time to viable designs by providing a set of flexible workflows to support the various stages of the design process using an integrated model hierarchy, ranging from the simple analytic descriptions to the highest fidelity, theory-based plasma and engineering modeling developed by the fusion and fission communities. These tools are expected to be needed for timely support of FPP design in the milestone program and in the FIRE collaboratives. The plasma simulation backbone of FREDA is IPS-FASTRAN with newly developed coupled Core-Edge Pedestal-SOL (CESOL) workflows, which is being extended to the far-SOL region up to the plasma facing components. FREDA incorporates the FERMI engineering modeling suite and will enable self-consistent evaluation of the thermal shields, limiters, blanket, magnets, and other surrounding structures with predictions of temperatures, erosion, dpa, activation, tritium generation and transport, creep, corrosion, material degradation, etc. Parametric generation of 3D CAD enables rapid iteration of component geometry in response to plasma and loading specifications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Effect of TiN coating on suppressing Ce-Fe interaction under irradiation

Advanced cladding is critical for fast reactors with the adequate thermal conductivity, mechanical stability and radiation tolerance of the cladding base material, corrosion resistance and high temperature coolant compatibility of the cladding surface, and chemical stability of the cladding inner wall against fuel cladding chemical interaction (FCCI). The preliminary results of recent ion irradiation studies of two diffusion-couple samples of cerium (Ce)/oxide-dispersion strengthened steel (ODS) and Ce/TiN/ODS, irradiated with 80 MeV xenon (Xe) ions to 100 displacements per atom (dpa) at 500°C, are summarized. Significant Ce-Fe interaction occurred in the Ce/ODS sample, and no noticeable Ce-Fe interaction was found in the Ce/TiN/ODS sample. It shows the effectiveness of 1-µm TiN diffusion barrier coated by the pulsed laser deposition on suppressing Ce-Fe interaction, a major contributor to FCCI in cladding. Here, density function theory (DFT) calculations of the impurity diffusivities of Ce and Fe within the Ti sublattice of TiN were performed to assist a mechanistic understanding of the experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗