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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
NASA NTRS1977

Upper bound on the equilibrium concentration of atomic H in solid H2

A phenomenological rate process theory has been developed for the production, storage and recombination of atomic H free radicals in a tritium-impregnated solid H2 at temperatures in the range of about 0.1 to 4 K. In this paper it is shown that the theory requires a stringent upper bound on the equilibrium concentration of trapped atomic H, namely that it cannot exceed about 0.125%, even if the temperature is reduced to an arbitrary low value and a very strong magnetic field is applied to the tritium-impregnated H2 solid.

Rosen, G.↗
DOE OSTI2023

Comparison of Designs of Hydrogen Isotope Separation Columns by Numerical Modeling

Mixtures of gas-phase hydrogen isotopologues (diatomic combinations of protium, deuterium, and tritium) can be separated using columns containing a solid such as palladium that reversibly absorbs hydrogen. A temperature-swing process can transport hydrogen into or out of a column by inducing temperature-dependent absorption or desorption reactions. Here, we consider two designs: a thermal cycling absorption process, which moves hydrogen back and forth between two columns, and a simulated moving bed (SMB), where columns are in a circular arrangement. We present a numerical mass and heat transport model of absorption columns for hydrogen isotope separation. It includes a detailed treatment of the absorption–desorption reaction for palladium. By comparing the isotope concentrations within the columns as a function of position and time, we observe that SMB can lead to sharper separations for a given number of thermal cycles by avoiding the remixing of isotopes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

First look at neutron emission shape characteristics of ignition hotspots at the National Ignition Facility (invited)

The nuclear imaging system has been capturing neutron images of inertial confinement fusion (ICF) driven implosions for over a decade at the National Ignition Facility. This imaging system has evolved from one to three nearly orthogonal lines-of-sight, allowing for the study of three-dimensional shape characteristics of ignition shots. Limited-view tomography algorithms help visualize the burning hotspot in 3D and assess neutron source geometry using Legendre mode parameters. With its neutron, gamma-ray, and x-ray image reconstruction capabilities, NIS has provided critical insight into mechanisms that have limited implosion performance, such as fill tube diameter for ignition-type targets. This comprehensive diagnostic suite opens a window into the shape characteristics of ignition shots and how symmetry affects ICF implosion performance. In more recent ignition shots, neutron yields have visibly increased. Analyzing the shape and size of the reconstructed neutron source has shown an expansion of the burn volume, which is indicative of more efficient alpha heating during the implosion process.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Deep Eutectic Solvents as Candidates for Lithium Isotope Enrichment

Nuclear fusion is a phenomenon that is well known within the nuclear physics community as a viable option for alternative energy as many natural gases and fossil fuels are phased out of commercial use. Deuterium and tritium fusion reactions are currently the leading candidates for nuclear fusion, with a major limiting factor being a means to produce tritium on an industrial scale. Lithium-6 is a well-known isotope that can produce tritium and helium following a fission reaction with a neutron. Unfortunately, the lithium-6 enrichment methods are limited to the COLEX process, which leaves behind an alarming amount of mercury waste as a potential environmental contaminant. Deep eutectic solvents are believed to be a potential alternative to lithium isotope separations due to the ease of generation, in addition to the minimum environmental waste generated when these solvents are employed. Previous studies have suggested that deep eutectic solvents are capable of separating lithium isotopes by utilizing a 2-thenoyltrifluoroacetone and trioctylphosphine oxide system that can biphasically react with a buffered solution containing lithium chloride. This system displays a separation factor of 1.068, which when compared to the 1.054 separation within the COLEX process, makes it a potential candidate for lithium-6/7 separation. Within this study, we investigate this system in comparison to two newly synthesized deep eutectic solvents and find that within these acetylacetone-based systems, little isotopic separation is observed. We investigate these systems both experimentally and computationally, showing the different lithium cation affinities, in addition to proposing how the electron-donating or -withdrawing nature can influence these systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Transport of Highly Volatile Gases Related to Noble Gases and Tritium in the MSRE

This study aims to comprehend mass transfer in the closed-loop circulation of highly volatile gases, including noble gases and tritium. We explore the impact of steady-state xenon-135 and tritium on the MSRE and reveal their isotopic distributions using online noble gas stripping of fuel salts. The MSRE was engineered to extract fission product gases from fuel salts and efficiently eliminate inert gases with the help of helium bubbles within a circulating fuel pump. These reactors introduce significant theoretical challenges in estimating interfacial area and mass transfer coefficients, crucial for modeling mass transfer processes. An essential component of our analysis is the mass transfer coefficient. These coefficients are important for understanding how radionuclides move during various phase transitions within a nuclear reactor. Xenon-135 and tritium are found in both liquid and gas phases within the reactor system. In the liquid phase, they dissolve in molten salts, while in the gas phase, they manifest as bubbles. These elements have significant adverse effects on reactor operation due to their strong neutron absorption properties, influencing both safety and performance. The Mole code, which predicts the behavior of chemical species under steady-state conditions, facilitates multiphysics coupling with Griffin to update species distributions and address inherent MSR safety.

Lee, Kyoung↗
DOE OSTI2024

Radioisotope Analysis of Wastewater from Livermore Site Retention Tanks by Gel Laboratory Gross Alpha, Gross Beta and Tritium Sampling Method

Lawrence Livermore National Laboratory discharged approximately 4.4% of the City of Livermore’s total wastewater in 2022 (LLNL’s Annual Site Environmental Report, Chapter 5, 2022). This volume includes wastewater from Sandia National Laboratories (SNL) and some process wastewater from Site 300. Due to the high volume and constituents of the discharge, LLNL works alongside the City of Livermore under permit #1250, requiring wastewater generated to be monitored and sampled in accordance with permit limits. Process wastewater, from buildings with the highest risk to sewer, is collected by wastewater retention tanks throughout the Livermore Site and sampled prior to discharge. Domestic wastewater directly discharges to sanitary sewer. To maintain permit requirements and ensure proper wastewater discharge practices, an internal wastewater audit was conducted during the summer of 2023. Current wastewater practices, regulatory knowledge and risk management across various Livermore Site buildings were evaluated. Workspaces connected to a wastewater retention tank and sanitary sewer drains were major focus areas. Data collected from walk-throughs prompted further evaluation as many practices were reported to be done based on historical usage. A table of concerns was created to showcase reasons for auditing and proceeding action. An analysis of current and historical retention tank usage throughout LLNL Livermore Site buildings with radioisotope results over a 5-year period from 2019 to 2024, was done to assess building trends and any significant changes throughout the 5-year period. Analytes evaluated were Gross Alpha, Gross Beta and Tritium (GABT) of eleven buildings at the Livermore Site, posing the highest risk to sanitary sewer for radioisotopes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Microstructural Effects of High Dose Helium Implantation in ErD 2

Metal hydrides can store hydrogen isotopes with high volumetric density. In metal tritides, tritium beta decay can result in accumulation of helium within the solid, in some cases exceeding 10 at.% helium after only 4 years of aging. Helium is insoluble in most materials, but often does not readily escape, and instead coalesces to form nanoscale bubbles when helium concentrations are near 1 at.%. Blistering or spallation often occurs at higher concentrations. Radioactive particles shed during this process present a potential safety hazard. This study investigates the effects of high helium concentrations on erbium deuteride (ErD 2 ), a non-radioactive surrogate material for erbium tritide (ErT 2 ). To simulate tritium decay in the surrogate, high doses of 120 keV helium ions were implanted into ErD 2 films at room temperature. Scanning and transmission electron microscopy indicated spherical helium bubble formation at a critical concentration of 1.5 at.% and bubble linkage leading to nanoscale crack formation at a concentration of 7.5 at.%. Additionally, crack propagation occurred through the nanocrack region, resulting in spallation extending from the implantation peak to the surface. Electron energy loss spectroscopy was utilized to confirm the presence of high-pressure helium in the nanocracks, suggesting that helium gas plays a predominant role in deformation. This work improves the overall understanding of helium behavior in ErD 2 by using modern characterization techniques to determine: the critical helium concentration required for bubble formation, the material failure mechanism at high concentration, and the nanoscale mechanisms responsible for material failure in helium implanted ErD 2 .

36 MATERIALS SCIENCE↗
DOE OSTI2024

Direct internal recycling fractions approaching unity

Direct internal recycling (DIR) refers to the process of recovering pure hydrogen isotopes (D/T) from helium and other impurities in the fusion plasma exhaust and directing them back to the fuel injection system. Increasing the exhaust fraction purified through DIR significantly reduces the size and cost of the tritium plant and provides additional benefits including reduced requirements for both the tritium startup inventory and tritium breeding ratio. Metal foil pumps (MFPs) are the dominant technology for this separation, relying on the concept of superpermeation. We recently demonstrated that PdCu foils operated at low temperature provide both exceptional flux and resilience to helium absorption as the DIR fraction is increased. Herein we design and demonstrate continuous and semi-batch DIR processes using PdCu MFPs. Under continuous processing, stable performance was observed for DIR fractions up to 92 %. In addition, we demonstrate a semi-batch process capable of extending the DIR fraction to unity (> 99.8 %). Under the experimental conditions described within a PdCu MFP area of ~22 m 2 would be sufficient to process the fusion exhaust with 92 % DIR fraction at expected flowrates of 100 Pa·m 3 ·s -1 for a future fusion power plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Halogen Removal from Tritium Gas Streams

Halogen impurities are poisonous and/or corrosive • Poison for catalysts and equipment • Causes pitting in most austenitic stainless steels • Minimized to prevent damage over time • Prone to appear in certain processes • Appear if certain materials degrade (i.e., elastomers, etc.) • Present as trace impurities in a large number of metals and zeolites (<200 ppm)

Angelette, Lucas M.↗
DOE OSTI2022

Mechanisms of shape transfer and preheating in indirect-drive double shell collisions

Implosions of Hohlraum-driven double shell targets as an alternative inertial confinement fusion concept are underway at the National Ignition Facility. The double shell system relies on a series of energy transfer processes starting from thermal x-ray absorption by the outer shell, followed by collisional transfer of kinetic energy to a heavy metal inner shell, and finally, conversion to the internal energy of the deuterium-tritium fuel. During each of these energy transfer stages, low-mode asymmetries can act to reduce the ideal transfer efficiency degrading double shell performance. Mechanisms, such as hard x-ray preheat from the Hohlraum, not only decrease the efficiency of kinetic energy transfer but may also be a source of low-mode asymmetry. In this article, we evaluate the shape transfer processes through the time of shell collision using two-dimensional integrated Hohlraum and capsule computations. We find that the dominant mode of the shape transfer is well described using a “radial impulse” model from the shape of the foam pressure reservoir. To evaluate the importance of preheat on inner shell shape, we also report on first measurements of Au L-shell preheat asymmetry in a double shell with a tungsten pusher. These measurements showed a 65% higher preheat velocity at the pole of the capsule relative to the equator. We also found that the experiments provided rigorous constraints by which to test the Hohlraum model settings that impact the amount and symmetry of Au L-shell preheat via the plasma conditions inside the outer cone Au bubble.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Approaching hydro-equivalent ignition in laser direct-drive via target design optimization using novel statistical modeling

Laser direct-drive offers significant advantages in terms of target simplicity, improved energy coupling, and large fuel masses over indirect drive. However, performance degradations from hydrodynamic and laser-plasma instabilities seeded and driven by the direct illumination pose limitations on the parameter space available for achieving ignition. In this paper, new design improvements are identified to forge a path forward for a hydro-equivalent ignition demonstration. The first is related to a new formulation of the statistical model (SM) used to accurately predict target performance directly from input parameters such as laser pulse shape and target specifications. This new SM formulation provides direct guidance on target dimensions and laser beam-to-target radius to achieve the highest fusion yield on the OMEGA laser. The second improvement comes from cooling the deuterium–tritium (DT) ice layer below the triple point right before shot time leading to lower DT vapor densities and higher convergence. Guided by these design improvements, a Bayesian optimization algorithm was used to design an implosion that is predicted to closely approach a Lawson triple product that hydrodynamically scales to ignition if equivalent laser–target coupling is achieved at laser energies typical of the National Ignition Facility.

Deuterium↗
DOE OSTI2025

Integrated Ammonia Capture and Exchange on Ion‐Exchanged 4A Zeolites

Ammonia poses a challenge in effluent gas streams due to its corrosive nature. However, in fusion settings tritiated ammonia can be formed, leading to both tritium loss in inventory and the generation of reactive species. Therefore, identifying pathways in which both tritium can be recovered, and the ammonia can be easily handled would be beneficial. One such way to do so would be to combine two useful techniques: ammonia sequestration and hydrogen exchange (ND 3 →NH 3 ). However, materials that can both adsorb ammonia and subsequently perform reactions on ammonia have not been well explored. Here, in this work, we present the development of ion-exchanged A-type zeolites to be utilized as a support material for platinum catalysts. In this way, the zeolite can adsorb ammonia and the platinum catalyst can facilitate hydrogen exchange allowing for bifunctional reactivity of the material to be achieved. A variety of elements were explored for their effect on A-type zeolites and resulted in an isotopic difference in the adsorption of ND 3 and NH 3 , noting the use of deuterium as a surrogate for tritium. Several platinum-impregnated zeolites were able to remove ND 3 from the gas stream, indicating that utilizing these materials in isotope recovery processes would improve accountability of these valuable hydrogen isotopes.

Koch, Christopher J. [Savannah River National Labo↗
DOE OSTI2022

Energetic particle optimization of quasi-axisymmetric stellarator equilibria

Abstract An important goal of stellarator optimization is to achieve good confinement of energetic particles such as, in the case of a reactor, alphas created by deuterium–tritium fusion. In this work, a fixed-boundary stellarator equilibrium was re-optimized for energetic particle confinement via a two-step process: first, by minimizing deviations from quasi-axisymmetry (QA) on a single flux surface near the mid-radius, and secondly by maintaining this improved QA while minimizing the analytical quantity Γ C , which represents the angle between magnetic flux surfaces and contours of J | | , the second adiabatic invariant. This was performed multiple times, resulting in a group of equilibria with significantly reduced energetic particle losses, as evaluated by Monte Carlo simulations of alpha particles in scaled-up versions of the equilibria. This is the first time that energetic particle losses in a QA stellarator have successfully been reduced by optimizing Γ C . The relationship between energetic particle losses and metrics such as QA error ( E q a ) and Γ C in this set of equilibria were examined via statistical methods and a nearly linear relationship between volume-averaged Γ C and prompt particle losses was found.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Tritiated Water Disposition Strategies

Tritium, a radioactive isotope of hydrogen, is typically confined within one or more confinement barriers. Tritium readily reacts with oxygen to form tritiated water or can exchange with protium or deuterium atoms in water to create tritiated water which is around 10,000 times more toxic than elemental tritium. Primary tritium confinement systems can create small volumes of highly tritiated water. Secondary confinement systems, like inert gloveboxes, will produce larger volumes of lower concentration tritiated water. Much larger volumes of tritiated water of very low tritium content are produced by tritium permeation/diffusion. Tritium from inside primary and secondary confinement barriers will diffuse from these systems into cooling water systems. Closed loop, recirculating cooling water systems will slowly accumulate tritium and create concentrations which exceed drinking water (or permitted discharge) concentrations. The cooling water systems designed as series of interconnected cooling loops that can slowly transfer tritium from one cooling loop to another cooling loop through heat exchangers. These large volumes of tritiated water may be above federal discharge limits and also lack convenient disposition pathways. This document will discuss tritiated water management strategies from the perspective of large volume, low tritium concentrations that exceed drinking water or permitted discharge limits. Applicable US regulatory requirements for a Molybdenum-99 (Mo-99) production facility concerning tritiated water are also cited. Tritiated water disposition options, based on typical waste minimization strategies will guide dispositions options. A Case Study example is presented to illustrate the pros and cons of several disposition options. The report does not recommend any particular disposition pathway but discusses strategies to be explored by a producer to guide selection of processes which best fits the needs of a particular waste stream.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Building a new multiphysics workflow in MOOSE: application to tritium migration, trapping and advection in TMAP8

Fusion devices are anticipated to produce and consume several kilograms of tritium per year. This rare fuel resource is both highly mobile and radioactive, making tracking inventories a priority for operation and safety. The fusion safety program at the Idaho National Laboratory has been developing the Tritium Migration and Analysis Program (TMAP), of which the latest version is a MOOSE-based application. TMAP8 is verified against its predecessors and possesses additional multi-dimensional tritium migration modeling capabilities. As we extend its capabilities towards both whole device (in multiple dimensions) and whole plant (with multiple components) simulations, the syntax of inputs must become compact, descriptive, compatible with quality assurance processes, and as error-proof as achievable. The new Physics system developed MOOSE can set up equations and instantiating them on plant components. The system permits the automatic definition of complex discretization with a consistency between object parameters achieved programmatically. The Physics system can currently instantiate the equations for heat conduction and Navier Stokes weakly compressible flow. In MOOSE-terms, it automates the definition of kernels, boundary conditions, and several core and helper materials and fields. As part of this effort, Physics classes were developed for tritium migration, trapping and advection within either a multi-dimensional Navier Stokes fluid dynamics simulation, or a 1D thermal hydraulics piping system. In this presentation, we will showcase the new syntax, its application to several verification and validation cases which were already studied using the classical TMAP8 syntax, and a demonstration of the new coupling capabilities for the migration of tritium into blanket coolant channels and the subsequent advection into the coolant loop.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Modeling ablator defects as a source of mix in high-performance implosions at the National Ignition Facility

Recent indirect drive inertial confinement fusion implosions on the National Ignition Facility (NIF) [Spaeth et al., Fusion Sci. Technol. 69, 25 (2016)] have crossed the threshold of ignition. However, performance has been variable due to several factors. One of the leading sources of variability is the quality of the high-density carbon (HDC) shells used as ablators in these experiments. In particular, these shells can have a number of defects that have been found to correlate with the appearance of ablator mix into the hot spot and a degradation in nuclear yield. These defects include pits on the ablator surface, voids in the ablator bulk, high-Z debris from the Hohlraum wall that adheres to the capsule surface, and finally the inherent granular micro-structure of the crystalline HDC itself. This paper summarizes high-resolution modeling of each of these mix sources in two recent high-performance NIF implosion experiments. The simulated impact from a range of individual capsule defects is found to be broadly consistent with the trends seen in experiment, lending credence to the modeling results and the details of the mixing process that they reveal. Interestingly, modeling of the micro-structure inherent to HDC shows that this perturbation source results in considerable mixing of the deuterium–tritium fuel with ablator material during the implosion. The reduction in fuel compression from this mix results in an approximately factor of two reduction in neutron yield in current implosions and emphasizes the importance of mitigating this significant performance degradation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Flowsheet Evaluation of Dissolving Used Nuclear Fuel in PUREX Solvent

This study explores the potential benefit to the used nuclear fuel actinide recovery flowsheet of a process simplification and establish the experimental work needed to develop the simplified flowsheet. The simplification is based on replacing acid dissolution of used nuclear fuel with dissolution in the tri-butyl phosphate solvent used in the industrially mature Plutonium Uranium Reduction Extraction flowsheet. Though characterized by considerable uncertainty, simplified flowsheets appear feasible and potentially offer significant reductions in process complexity, nitrate inventory, secondary liquid effluent generation and plant footprint. Initial development of the technology should focus on key fission product and actinide dissolution as a function of process conditions and understanding the fundamental chemistry of tri-butyl phosphate solvent chemistry. Development of ancillary technologies for minor actinide separations and fuel pretreatment for tritium management is also recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Chemical Thermodynamic Modeling of Molten Salts to Support Off-Gas Abatement Systems

The reprocessing of used nuclear fuel by any means will liberate gaseous fission products, such as hydrogen ( 3 H), carbon ( 14 C), noble gases ( 85 Kr), and halogens ( 129 I), from the irradiated fuel. These elements will distribute through chemical processing operations and partition into process off-gas streams, and the specific volatile release fractions will be dictated by the chemical and physical properties of the system. A recent assessment found that there were significant knowledge gaps regarding the release quantities of volatile radionuclides from individual unit operations. These knowledge gaps limited the ability to determine what dedicated off-gas treatment technologies could be required for electrochemical-based reprocessing facilities. Unfortunately, experimental efforts to quantify release fractions are limited by the challenges associated with performing experiments using irradiated fuel. This report documents preliminary thermodynamic predictions of iodine and tritium release from chloride-based molten salts as part of an effort to better direct resources toward those experiments (both simulant and irradiated) that will be of the greatest impact. It was predicted that less than 0.5% of tritium was expected to be released and that nearly all of that amount would be released as H 2 . Thermochemical data for hydrogen (H 2 ) are of high fidelity, and no additional validation is recommended. Less than 0.05% of iodine was predicted to be released, with the primary volatile species being Cs 2 I 2 . Unlike the tritium predictions, the data underpinning the iodine release predictions are of low quality. It is recommended that a limited experimental program be dedicated to expanding the current physical property and thermodynamic property data for iodine in electrochemical processing and molten salt conditions, including vapor pressure measurements for the dimerized species predicted to comprise the majority of iodine release. Validating and improving the thermodynamic properties used in predictive modeling can reduce the need for expensive testing with irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗