Results for “surface reactor” — Page 80

DOE OSTI2021

Automated Cyber Security Testing Platform for Industrial Control Systems

Nuclear Power Plants (NPPs) are a complex system of coupled physics controlled by a network of Programmable Logic Controllers (PLCs). These PLCs communicate process data across the network to coordinate control actions with each other and inform the operators of process variables and control decisions. Networking the PLCs allows more effective process control and provides the operator more information which results in more efficient plant operation. This interconnectivity creates new security issues, as operators have more access to the plant controls, so will bad actors. As plant networks become more digitized and encompass more sophisticated controllers, the network surface exposed to cyber interference grows. Understanding the dynamics of these coupled systems of physics, control logic, and network communications is critical to their protection. The research into the cybersecurity of the Operational Technologies of NPPs is developing and requires a platform that can allow high fidelity physics simulations to interact with digital networks of controllers. This will require three main components: a network simulation environment, a physics simulator, and virtual PLCs (vPLC) that represent typical industry hardware. A platform that incorporates these three components to provide the most accurate representation of actual NPP networks and controllers is developed in this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Deformation band corrosion as a result of applied stress and cold-work in 304 stainless steel in simulated pressurized water reactor primary water conditions

Stainless steels like 304SS are commonly used in pressurized water reactor primary water environments due to their excellent resistance to stress corrosion cracking (SCC). However, SCC susceptibility has been observed in cold-worked 304SS. Cold-rolling introduces a high density of defects, such as deformation bands or dislocations, into the microstructure; applied stress, even if macroscopically below the yield stress, can result in regions with high localized stresses. Cold-worked and non-cold-worked samples were exposed to 360 °C simulated PWR primary water for 1500 h with and without active loading. Corrosion susceptibility of dislocations and deformation bands induced by cold-rolling and/or applied stress was studied using high-resolution (analytical) transmission electron microscopy. Deformation bands were generated in all specimens examined in the study. Penetrative oxidation was observed in some deformation bands in each specimen. The depth of penetrative deformation band oxidation frequently exceeded the depth of grain boundary oxidation in all specimens. However, deformation band oxides generally were narrowly confined to the deformation band plane. Highly Ni-rich regions were observed beyond all oxidation fronts. Surface oxides were thicker in the specimens exposed under active loading. In conclusion, the implications of these findings at surfaces for deformation bands intersecting stress corrosion crack paths are discussed.

EFTEM↗

DOE OSTI2021

Measurement of Fission Product Concentration Profiles in AGR-3/4 TRISO Fuel Graphitic Matrix and Nuclear Graphites

The third Advanced Gas Reactor (AGR) irradiation experiment, AGR-3/4, was designed to investigate the migration of fission products in fuel compact graphitic matrix and reactor graphite components. Using destructive methods, radial fission product concentration profiles were measured for gamma-emitting fission products (e.g., Ag-110m, Cs-134, and Eu-154) and beta-emitting Sr-90 in irradiated graphitic and graphite components from six different AGR-3/4 irradiation capsules. These new measured concentration profiles can now be compared to non-destructive measurements and fission product transport simulations and will be used to derive new diffusivities and sorptivities to support refinement of fission product transport models and high-temperature gas-cooled reactor (HTGR) source-term analyses. Each capsule in the AGR-3/4 experiment had four fuel compacts in the middle of two concentric rings of graphitic matrix material, PCEA graphite, or IG-110 graphite. In addition to the approximately 1898 tristructural isotropic (TRISO) coated particles in each compact, there were 20 designed to fail (DTF) particles coated only in pyrocarbon so that they released fission products into the surrounding cylindrical rings of carbonaceous materials. Destructive sampling of the rings involved machining/milling material from around the circumference of the rings, collecting that material, and performing radiochemical analyses on it. Milling operations were performed in multiple steps or segments, and each segment was generally 0.508 mm (0.020 in) thick. Knowing the radial position at which each segment was milled, the volume of the milled material at each segment, and the fission product content in each segment, the radial fission product concentrations were constructed for select isotopes in each ring. Ag-110m profiles had the most variation. Some profiles were peaked at an inner or outer surface. Some were peaked at the middle of the ring wall thickness. Some increased radially outward, and some decreased radially outward. These types of variations and the fact that the measured profiles do not generally compare favorably with the transport model employed for AGR-3/4 may adversely impact the ability to extract reasonable transport parameters for this isotope. In many cases, the Cs-134 profiles decreased somewhat linearly in the outward radial direction, and in cursory comparisons, the shapes of these profiles appeared similar to those from model predictions. The step changes in concentration across the inner-outer ring gap were generally consistent with the model predictions as well. In some cases, there were local maxima in concentration at the outer surface of the rings. This suggests that fission products could have transported in the small gaps between the inner ring and the outer ring and between the outer ring and the sink ring such that some portion of a given fission product can bypass diffusion through the ring itself. The analysis of the small nubs on the outer surfaces of some of the outer rings revealed fission product concentrations in the nubs that were often higher than in the outermost segments of the rings. This further supports the hypothesis that short-circuit, gap transport occurred, causing relatively high surface concentrations on the outer surfaces of the rings. Eu-154 and Sr-90 profiles tend to have very similar shapes, suggesting that they transport via the same mechanisms. The observed profiles were indicative of a transport process where the isotopes are sorbed on the inner surface of the ring, but diffusion into the ring from that surface is quite slow. Some elevated concentrations of Sr-90 (relative to Eu-154) on the outer surface of a ring suggested that rapid, gas-gap transport of gaseous precursor Kr-90 and volatile Rb-90 could have occurred prior to their decaying to Sr-90. Overall, the Eu 154 and Sr-90 profiles were still very similar, which indicates that the transport of short-lived Sr-90 precursors is not a major effect. In some capsules, the qualitative Sr-90 behavior across the ring gaps was consistent with the model (using the available legacy Sr-90 transport parameters), but in other capsules the model was inconsistent with the measurements and seems to underestimate the amount of Sr-90 in the outer rings. The total ring Sr-90 inventories were estimated for all the rings that were subject to physical sampling. These results will be used to adjust the predicted particle and/or compact releases used in the AGR-3/4 fission product transport model. Given the different irradiation temperatures among the capsules and the rings, it was not possible to discern fundamental differences in the transport of isotopes within the different carbon materials, i.e., graphitic matrix, IG-110, or PCEA. It may be possible to do this in the course of determining transport from the concentration profiles in future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Low-cost desalination of seawater and hypersaline brine using nanophotonics enhanced solar energy membrane distillation

A stand-alone small-scale nanophotonics enhanced solar membrane distillation (NESMD) testbed was designed, developed, and tested for desalinating seawater and high salinity feedwaters. This NESMD system can take almost any source water and turn it into clean water with sunlight as the only energy source. The NESMD technology applies nanophotonic coating materials on a commercial hydrophobic polypropylene support of a polytetrafluoroethylene (PTFE) membrane surface. The photothermal coating serves as a solar thermal collector, absorbs solar energy, and generates highly localized heat on the membrane, while the rest of the membrane performs membrane distillation functions. The presented NESMD system is equipped with an internal heat recovery system with no need for external electricity and no additional water for cooling. The NESMD testbed is installed and tested at Rice University campus (29.7174° N, 95.4018° W). In this study, real seawater from Galveston Bay, Texas, U.S. and high salinity simulated feedwaters (total dissolved solids (TDS) of 113 200–200 000 PPM) have been tested for long-term testing under the weather conditions of Houston, Texas. Here, the field testing results showed stable desalination performance in consecutive 5–8 hour operation cycles, with a TDS removal of ≥99.5% in all experiments. An average daily membrane flux of ≥0.75 L m –2 h –1 was achieved at a solar intensity close to 1 kW m –2 without an external heat exchanger. Further investigations and improvements are required to enhance the performance of the reactor since it is still a new promising technology.

14 SOLAR ENERGY↗

NASA NTRS2005

Closed-Cycle Engine Program Used to Study Brayton Power Conversion

One form of power conversion under consideration in NASA Glenn Research Center's Thermal Energy Conversion Branch is the closed-Brayton-cycle engine. In the tens-of-kilowatts to multimegawatt class, the Brayton engine lends itself to potential space nuclear power applications such as electric propulsion or surface power. The Thermal Energy Conversion Branch has most recently concentrated its Brayton studies on electric propulsion for Prometheus. One piece of software used for evaluating such designs over a limited tradeoff space has been the Closed Cycle Engine Program (CCEP). The CCEP originated in the mid-1980s from a Fortran aircraft engine code known as the Navy/NASA Engine Program (NNEP). Components such as a solar collector, heat exchangers, ducting, a pumped-loop radiator, a nuclear heat source, and radial turbomachinery were added to NNEP, transforming it into a high-fidelity design and performance tool for closed-Brayton-cycle power conversion and heat rejection. CCEP was used in the 1990s in conjunction with the Solar Dynamic Ground Test Demonstration conducted at Glenn. Over the past year, updates were made to CCEP to adapt it for an electric propulsion application. The pumped-loop radiator coolant can now be n-heptane, water, or sodium-potassium (NaK); liquid-metal pump design tables were added to accommodate the NaK fluid. For the reactor and shield, a user can now elect to calculate a higher fidelity mass estimate. In addition, helium-xenon working-fluid properties were recalculated and updated.

Johnson, Paul K.↗

DOE OSTI2021

Measurement of radiolytic hydrogen generation and impact of drying treatments on reactor exposed and surrogate aluminum materials

Technical challenges associated with dry storage of Aluminum-clad Spent Nuclear Fuel (ASNF) include a need to better understand the potential extent of gaseous molecular hydrogen (H 2 ) production through radiolytic degradation of the aluminum (oxy)hydroxide films present on ASNF. The characterization of radiolytic gas generation from ASNF (oxy)hydroxide layers has been identified as a key knowledge gap which poses a technical challenge to the long-term storage of ASNF. Task 2 of the action plan for the extended (>50 years) dry storage of ASNF addresses this gap. Previous radiolysis studies performed under Task 2 established baseline estimates of H 2 generation rates from the attendant hydrated oxides. The results confirmed that net radiolytic H 2 production has a dependency on absorbed gamma dose, as well as relative humidity and cover gas composition (air, nitrogen, and argon). Further experiments revealed that the physisorbed water on the samples may significantly impact the radiolytic H 2 yield. This phenomenon complicates the determination of H 2 generation rates for hydrated oxides, which may lead to inaccurate modeling predictions of the long-term H 2 yields in sealed storage systems containing ASNF, particularly when compared to spent fuel casks which have undergone some drying process intended to remove physically- and chemically-bound water. This report describes the testing methods utilized and the hydrogen generation results obtained in an investigation of the effects of gamma irradiation on aluminum materials for a variety of (oxy)hydroxide surface compositions and drying conditions. The testing methods included small-area aluminum material testing in ampules, and large-area aluminum material testing in steel vessels. Test material preparation, irradiation, and radiolytic H 2 measurement methods are summarized. The measured H 2 concentrations, which reflect the variable initial hydrated inventory and drying treatments, are compared to one another as well as to previously published data, to identify the primary factors affecting radiolytic H 2 generation rates and potential equilibrium H 2 concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRSDate not supplied

Quantitation of Trace Water in ISS Atmosphere Samples Recovered from CO2-Removal Systems

As crewed spaceflight continues to push the limits of exploration, instruments and hardware used to maintain and monitor crew and vehicle health must become more integrated, reliable, and efficient. By combining a carbon dioxide (CO2) removal system with a resource recovery system like a Sabatier or Bosch reactor, the waste gas from the removal system can be converted into valuable resources that help reduce reliance on ground resupply. Due to the high-pressure nature of the Sabatier and Bosch reactions, the vent gas from most CO2 removal systems will need to be compressed before it can be processed. In order to prevent condensation in the compressor of the recovery system, it is critical to ensure that the vent stream from the CO2 removal system does not contain an excess of water. There are currently three different CO2 removal systems that are being developed as candidates for exploration missions. Each of these systems will be evaluated as a technology demonstration on ISS, and vent gas samples will be collected in Summa sampling canisters fitted with Entech valves. Due to the need to measure low levels of water in these samples, it was not possible to use the current ISS miniature grab sample containers (GSCs) to collect these samples. The Summa containers, which were used to collect archive samples during Shuttle missions, have a different surface treatment that makes it possible to measure the low levels of water in samples collected in these containers. This paper will provide some background on the CO2 removal systems being evaluated for exploration missions as well as results from ground-based testing of the current miniature GSCs and Summa canisters. Details on the development of the analytical method and the plan for preparing the canisters that will be used for in-flight sampling are also discussed.

Steven W Beck↗

DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗

DOE OSTI2020

Chemical Modeling of the Capture of Radioiodine Species

Iodine is a common fission product resulting from the transmutation of uranium fuel in nuclear power reactors. The short-lived radioactive isotope, 131 I, with a half-life of around eight days, would be a particular health concern if released into the environment. Capture materials, such as activated carbon, are used routinely in nuclear power plants and medical isotope production facilities to remove this hazard during normal operations to reduce emissions to acceptable levels. Iodine, unlike other common fission products such as noble gases, is challenging because of the wide range of molecular forms and oxidation states (from +VII in IF 7 to -I in iodide) that it displays, which can interconvert both in the gas phase and in the solution phase for example. The work in this project is focused on three computational tasks. In the first, we consider the binding of iodine-containing molecules of varying oxidation states to the major component of activated carbon – pristine graphene-like sheets. The second task focused on molecular binding and possible chemical transformations resulting from the reaction of molecular iodine with chemical functional groups on the edge of the graphene sheets or surface defects in the sheet itself. The final task was associated with radiolytic breakdown of 131 I to 131 Xe in iodine containing molecules and investigated the resulting molecular fragmentation process both in the gas phase and on the graphene-like surface. For each of the tasks, the intent was to learn about the thermodynamics and kinetics of the chemical transformations and use the calculations to provide parameters for input into kinetic models describing these processes for interpretation of experimental data in future work.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Selective Methane Oxidation to Methanol on ZnO/Cu 2 O/Cu(111) Catalysts: Multiple Site-Dependent Behaviors

Due to the abundance of natural gas in our planet, a major goal is to achieve a direct methane to methanol conversion at medium to low temperatures using mixtures of methane and oxygen. Here, we report an efficient catalyst, ZnO/Cu 2 O/Cu(111), for this process investigated using a combination of reactor testing, scanning tunneling microscopy, ambient-pressure x-ray photoemission spectroscopy, density functional calculations, and kinetic Monte Carlo simulations. The catalyst is capable of methane activation at room temperature and transforms mixtures of methane and oxygen to methanol at 450 K with a selectivity of ~30%. This performance is not seen for other heterogenous catalysts which usually require the addition of water to enable a significant conversion of methane to methanol. The unique coarse structure of the ZnO islands supported on a Cu 2 O/Cu(111) substrate provides a collection of multiple centers that display different catalytic activity during the reaction. ZnO-Cu 2 O step sites are active centers for methanol synthesis when exposed to CH 4 and O 2 due to an effective O-O bond dissociation, which enables a methane-to-methanol conversion with a reasonable selectivity. Upon addition of water, the defected O-rich ZnO sites, introduced by Zn vacancies, show superior behavior towards methane conversion and enhance the overall methanol selectivity to over 80%. Thus, in this case, the surface sites involved in a direct CH 4 → to CH 3 OH conversion are different from those engaged in methanol formation without water. Furthermore, the identification of the site-dependent behavior of ZnO/Cu 2 O/Cu(111) opens a design strategy for guiding efficient methane reforming with high methanol selectivity.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Optimization of lithium vapor box divertor evaporator location on NSTX-U using SOLPS-ITER

Commercial fusion reactors will be faced with extremely high divertor target heat fluxes that will require mitigation. Simulations of detachment in an NSTX-U scenario projected to have 92 MW/m 2 unmitigated peak target heat flux are presented, which reaches sub-10 MW/m 2 target heat flux using a highly dissipating lithium vapor box divertor design. The lithium vapor box is a detached divertor design which employs lithium vapor evaporation and condensation to contain lithium below the X-point. Previous SOLPS modeling has indicated a lithium vapor box can reduce the heat flux down to 10 MW/m 2 via simultaneous evaporation from the Private Flux Region (PFR) and the Common Flux Region (CFR) sides of the vapor box. It is found here that PFR evaporation has improved access to the separatrix leading to significantly more efficient power dissipation than CFR evaporation. Simulations of target evaporation with an evaporation distribution that is self-consistent with the temperature of a Capillary Porous System with Fast flowing liquid lithium (CPSF) could reach n Li / n e ~ 0.025-0.030 at the Last Closed Flux Surface (LCFS) depending on the liquid metal flow speeds and lithium sputtering yield, while PFR-side evaporation can reach acceptable heat fluxes with n Li / n e ~ 0.038 at the LCFS. However, PFR evaporator performance can be improved if the target is allowed to be hot enough such that it reflects lithium, reaching n Li / n e ~ 0.028 and reducing required lithium evaporation. Ultimately PFR evaporation and target evaporation are found to have similar ability to produce acceptable heat flux solutions with minimal upstream concentration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRSDate not supplied

Quantitation of Trace Water in ISS Atmosphere Samples Recovered from CO2-Removal Systems

As crewed spaceflight continues to push the limits of exploration, instruments and hardware used to maintain and monitor crew and vehicle health, must become more integrated, reliable, and efficient. By combining a carbon dioxide (CO2) removal system with a resource recovery system like a Sabatier or Bosch reactor, the waste gas from the removal system can be converted into valuable resources that help reduce reliance on ground resupply. Due to the high-pressure nature of the Sabatier and Bosch reactions, the vent gas from most CO2 removal systems will need to be compressed before it can be processed. In order to prevent condensation in the compressor of the recovery system, it is critical to ensure that the vent stream from the CO2 removal system does not contain an excess of water. There are currently two different CO2 removal systems that are being developed as candidates for exploration missions. Each of these systems will be evaluated as a technology demonstration on ISS, and vent gas samples will be collected in Summa sampling canisters fitted with Entech valves for ground analysis. Due to the need to measure low levels of water in these samples, it was not possible to use the current ISS mini-grab sample containers (m-GSC) to collect these samples. The Summa containers, which were used to collect archive samples during Shuttle missions, have a different surface treatment that makes it possible to measure the low levels of water in samples collected in these containers. This paper will provide some background on the CO2 removal systems being evaluated for exploration missions as well as results from ground-based testing of the Summa canisters. Details on the development of the analytical method and the plan for preparing the canisters that will be used for in-flight sampling are also discussed.

William M. King↗

DOE OSTI2022

Challenges and Opportunities for the Development of Commercial Maritime Surface Vessel Nuclear Propulsion (CMNP)

Maritime transport and trade are currently responsible for producing nearly 1 billion metric tons of CO2 emissions annually, representing 2.5% of global greenhouse gas emissions. While maritime shipping is acknowledged as the most efficient carrier of goods in terms of its costs and carbon intensity, the sector relies overwhelmingly on the use of heavy fuel oil (HFO). This fuel source currently is unmatched in terms of its global availability, safety, security of supply, price, and energy density. Combined, these characteristics are responsible for the outsized contribution of the sector to the global economy: roughly 90% of all transported goods spend time at sea. Commercial maritime nuclear propulsion provides opportunities to transition maritime trade to clean energy but also comes with challenges. Both are covered in this report based on 60+ interviews with professionals from the maritime and nuclear energy sectors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

CFD modeling of turbulent air flow in self-heated gyroid TPMS structures: Thermal-hydraulic performance and validation

The application of mathematically derived geometries, such as triply periodic minimal surface (TPMS) lattices, has garnered significant interest across various fields, including the nuclear sector, due to their superior thermal-hydraulic characteristics for heat transfer compared to traditional plain or finned tubes. Here, this study validates a computational fluid dynamics (CFD) model, evaluates different turbulence models and CFD model settings, and performs uncertainty quantification to provide a comprehensive analysis. Despite extensive research on CFD modeling of TPMS lattices, such as gyroid and diamond geometries, there is a notable lack of publicly available literature providing comprehensive details on numerical analysis aspects, including convergence and methodological best practices. This study embarks on a benchmark analysis of a gyroid geometry to evaluate its thermal-hydraulic performance under turbulent flow conditions and scrutinize various CFD model configurations. The main contributions of this work include validating the CFD model, assessing and comparing different turbulence models, and enhancing pressure drop and temperature prediction capabilities. The results aim to support the development of methodologies needed to benchmark and enhance numerical analysis techniques for TPMS lattices. This work seeks to complement the existing body of knowledge, support the development of TPMS reactor concepts, and improve best practices for CFD modeling of TPMS lattices, ultimately advancing methodologies to support future applications in this domain.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Process design tools and techno-economic analysis for capacitive deionization

Capacitive deionization (CDI) devices use cyclical electrosorption on porous electrode surfaces to achieve water desalination. Process modeling and design of CDI systems requires accurate treatment of the coupling among input electrical forcing, input flow rates, and system responses including salt removal dynamics, water recovery, energy storage, and dissipation. Techno-economic analyses of CDI further require a method to calculate and compare between a produced commodity (e.g. desalted water) versus capital and operational costs of the system. In this work, we demonstrate a new modeling and analysis tool for CDI developed as an installable Matlab program that allows direct numerical simulation of CDI dynamics and calculation of key performance and cost parameters. The program is provided for free and is used to run open-source Simulink models. The Simulink environment sends information to the program and allows for a drag and drop design space where users can connect CDI cells to relevant periphery blocks such as grid energy, battery, solar panel, waste disposal, and maintenance/labor cost streams. The program allows for simulation of arbitrary current forcing and arbitrary flow rate forcing of one or more CDI cells. We employ validated well-mixed reactor formulations together with a non-linear circuit model formulation that can accommodate a variety of electric double layer sub-models (e.g. for charge efficiency). The program includes a graphical user interface (GUI) to specify CDI plant parameters, specify operating conditions, run individual tests or parameter batch-mode simulations, and plot relevant results. The techno-economic models convert among dimensional streams of species (e.g. feed, desalted water, and brine), energy, and cost and enable a variety of economic estimates including levelized water costs.

42 ENGINEERING↗

DOE OSTI2020

Deuterium retention characteristics in Li film by coating and during flowing liquid Li limiter operation in experimental advanced superconducting tokamak

Lithium (Li) is a promising low-Z material for particle recycling and impurity control to improve plasma performance in fusion devices. In the experimental advanced superconducting tokamak (EAST), Li coating has become a routine method for wall conditioning, and a flowing liquid Li (FLiLi) limiter has been successfully tested several times. Deuterium retention characteristics in the Li film coated on the international thermonuclear experimental reactor-like tungsten divertor and FLiLi during plasma discharges, which is important for the utilization of Li in future fusion devices, were investigated in EAST. It is found that the absorption of the fuel particles by Li coatings decreases gradually, and recycling gradually increases over a series of discharges. The maximum net amount of deuterium retained reached ~0.8 g, corresponding to 12% deuterium in the Li. Furthermore, this corresponds to a whole day's worth of shots with a total of 87 plasma discharges and a total of ~640s plasma time after 11.75 g Li was deposited. Compared to the Li coating, it is shown that FLiLi continuously traps fuel particles and achieves a higher deuterium retention ratio over both the short- and long-term, leading to lower recycling. Meanwhile, it is also observed that the fuel particle retention ratio increases when FLiLi is closer to the plasma. This result is likely due to higher plasma heating power and limiter temperature, which cause an increased Li efflux from FLiLi due to a higher limiter temperature, which then redeposits on the other plasma-facing surfaces and increases fuel particle absorption. It is estimated that ≥ 80% of the retained D particles are captured by the continual renewal of the Li redeposition film during the FLiLi operation. This investigation would also be useful for D/T retention in future fusion devices if Li is used as a plasma-facing component.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Influence of Ordered Mesoporous Oxides in Plasma-Assisted Ammonia Synthesis

Widespread implementation of dielectric barrier discharge (DBD)-assisted NH 3 synthesis, a nascent technology operating under sustainable, ambient conditions, is hindered by low energy yields due to, in part, poor fundamental understanding. Porous oxides used to support metal nanoparticle catalysts have shown significant energy yield contributions for DBD-assisted NH 3 synthesis even without metal. Using an AC-powered, coaxial, single-stage reactor at 16 kV with equimolar (N 2 /H 2 ) feed, we measured NH 3 synthesis rates in the presence of different nonordered oxides, ordered SiO 2 structures (SBA-15 and MCM-41), and ordered Al-incorporated analogues (γ-Al 2 O 3 -coated with varying Al-loadings and Al-substitution, respectively: Al 2 O 3 -SBA-15 and Al-MCM-41). We systematically quantified NH 3 energy yield dependence on pore structures and material identities (i.e., ordered pores and Al incorporation) known to facilitate higher DBD-assisted NH 3 synthesis rates. SBA-15 displayed a higher steady-state energy yield than MCM-41, indicating that framework type is a crucial factor, with both ordered porous systems outperforming fumed SiO 2 . 10 wt % Al maximized in situ NH 3 uptake among the various Al loadings, exhibiting a higher steady-state energy yield and similar power to SBA-15. However, Al-MCM-41 had a similar steady-state energy yield and lower power than MCM-41, likely due to the extended γ-Al 2 O 3 surface that has a dielectric constant higher than that of SiO 2 . Both Al-incorporated analogues benefit from surface acid sites that can adsorb NH 3 in situ, resulting in higher overall NH 3 energy yields than that of their parent ordered SiO 2 . Al 2 O 3 -SBA-15 shielded more NH 3 than Al-MCM-41, likely due to a higher acid site density than the acid site identity. Furthermore, Al incorporation via γ-Al 2 O 3 coating more successfully improves the NH 3 energy yield; together with the high-performing ordered framework, these analogues are potential metal catalyst supports with promising energy yields for DBD-assisted synthesis of NH 3 and other chemicals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗