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DOE OSTI2021

Nanoscale biochemical sample preparation and analysis

Provided herein are methods and systems for biochemical analysis, including compositions and methods for processing and analysis of small cell populations and biological samples (e.g., a robotically controlled chip-based nanodroplet platform). In particular aspects, the methods described herein can reduce total processing volumes from conventional volumes to nanoliter volumes within a single reactor vessel (e.g., within a single droplet reactor) while minimizing losses, such as due to sample evaporation.

Kelly, Ryan T.↗

DOE OSTI2020

Irradiated Microstructure of Zorita Materials

Reactor baffle plate materials were harvested from the decommissioned Zorita nuclear power reactor after 38 years of service and were characterized with transmission electron microscope. The samples were Type 304 stainless steels and had been irradiated to 0.05, 16.9, ad 48.6 dpa. Faulted dislocation loops, voids, and irradiation-induced precipitates were observed in this material. While dislocation loops were the dominant irradiation defects for all samples, small voids and precipitates were also observed at 16.9 and 48.6 dpa, but not at 0.05 dpa. The size of dislocation loops increased with irradiation dose and saturated below 16.9 dpa. The loop density continued to increase between 16.9 and 48.6 dpa. For voids, both the density and mean size were small, and the volume swelling was negligibly small even at 48.6 dpa. In addition, carbides or G-phase precipitates were detected at 16.9 and 48.6 dpa, and a complex dose dependence was observed. A range of irradiation hardening was estimated at each dose, and the resulting yield strengths were consistent with the previous tensile results measured from the Halden and BOR-60 specimens.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Advancing Tritium Self-Sufficiency in Fusion Power Plants: Insights from the BABY Experiment

Abstract In the pursuit of fusion power, achieving tritium self-sufficiency stands as a pivotal challenge.Tritium breeding within molten salts is a critical aspect of next-generation fusion reactors, yet experimental measurements of \gls{tbr} have remained elusive.Here we present the results of the \gls{baby} experiment, which represents a pioneering effort in tritium research by utilizing high-energy (\SI{14}{\mega\electronvolt}) neutron irradiation of molten salts, a departure from conventional low-energy neutron approaches. Using a small-scale (\SI{100}{\milli\litre}) molten salt tritium breeding setup, we not only simulated, but also directly measured a \gls{tbr} \textcolor{red}{(\num{3.57e-4})}. This innovative approach provides crucial experimental validation, offering insights unattainable through simulation alone.Moreover, our findings reveal a surprising outcome: tritium was predominantly collected as HT, contrary to the expected TF. This underscores the complexity of tritium behavior in molten salts, highlighting the need for further investigation.This work lays the foundation for a more sophisticated experimental setup, including increasing the volume of the breeder, enhancing neutron detection, and refining tritium collection systems.Such improvements are crucial for advancing our understanding of fusion reactor feasibility and paving the way for future experiments.

Delaporte-Mathurin, Rémi (ORCID:0000000310648882)↗

NASA NTRS2001

The Cool Flames Experiment

A space-based experiment is currently under development to study diffusion-controlled, gas-phase, low temperature oxidation reactions, cool flames and auto-ignition in an unstirred, static reactor. At Earth's gravity (1g), natural convection due to self-heating during the course of slow reaction dominates diffusive transport and produces spatio-temporal variations in the thermal and thus species concentration profiles via the Arrhenius temperature dependence of the reaction rates. Natural convection is important in all terrestrial cool flame and auto-ignition studies, except for select low pressure, highly dilute (small temperature excess) studies in small vessels (i.e., small Rayleigh number). On Earth, natural convection occurs when the Rayleigh number (Ra) exceeds a critical value of approximately 600. Typical values of the Ra, associated with cool flames and auto-ignitions, range from 104-105 (or larger), a regime where both natural convection and conduction heat transport are important. When natural convection occurs, it alters the temperature, hydrodynamic, and species concentration fields, thus generating a multi-dimensional field that is extremely difficult, if not impossible, to be modeled analytically. This point has been emphasized recently by Kagan and co-workers who have shown that explosion limits can shift depending on the characteristic length scale associated with the natural convection. Moreover, natural convection in unstirred reactors is never "sufficiently strong to generate a spatially uniform temperature distribution throughout the reacting gas." Thus, an unstirred, nonisothermal reaction on Earth does not reduce to that generated in a mechanically, well-stirred system. Interestingly, however, thermal ignition theories and thermokinetic models neglect natural convection and assume a heat transfer correlation of the form: q=h(S/V)(T(bar) - Tw) where q is the heat loss per unit volume, h is the heat transfer coefficient, S/V is the surface to volume ratio, and (T(bar) - Tw ) is the spatially averaged temperature excess. This Newtonian form has been validated in spatially-uniform, well-stirred reactors, provided the effective heat transfer coefficient associated with the unsteady process is properly evaluated. Unfortunately, it is not a valid assumption for spatially-nonuniform temperature distributions induced by natural convection in unstirred reactors. "This is why the analysis of such a system is so difficult." Historically, the complexities associated with natural convection were perhaps recognized as early as 1938 when thermal ignition theory was first developed. In the 1955 text "Diffusion and Heat Exchange in Chemical Kinetics", Frank-Kamenetskii recognized that "the purely conductive theory can be applied at sufficiently low pressure and small dimensions of the vessel when the influence of natural convection can be disregarded." This was reiterated by Tyler in 1966 and further emphasized by Barnard and Harwood in 1974. Specifically, they state: "It is generally assumed that heat losses are purely conductive. While this may be valid for certain low pressure slow combustion regimes, it is unlikely to be true for the cool flame and ignition regimes." While this statement is true for terrestrial experiments, the purely conductive heat transport assumption is valid at microgravity (mu-g). Specifically, buoyant complexities are suppressed at mu-g and the reaction-diffusion structure associated with low temperature oxidation reactions, cool flames and auto-ignitions can be studied. Without natural convection, the system is simpler, does not require determination of the effective heat transfer coefficient, and is a testbed for analytic and numerical models that assume pure diffusive transport. In addition, mu-g experiments will provide baseline data that will improve our understanding of the effects of natural convection on Earth.

Pearlman, Howard↗

DOE OSTI2020

ZPPR-15 Criticality Measurements

A reference critical configuration was established for each phase of the ZPPR-15 experimental program. The reference critical configuration for each phase corresponds to a configuration in which the measured excess reactivity was adjusted to a standard set of reporting conditions with an average reactor temperature of 293 K, a zero interface gap and all operational control rods parked 30 inches from the axial centerline of the core. The ZPPR staff published the measured excess reactivity and the statistical uncertainty in the measurement. In reality, the published uncertainty for each reference critical configuration is only a small component of the total uncertainty for that configuration. Additional uncertainties related measurement technique, configuration geometry and material compositions exist. The geometry and material composition uncertainties are the dominant uncertainties for ZPPR-15. A full uncertainty analysis was performed for each ZPPR-15 reference critical configuration. The significant uncertainties related to measurement technique, configuration geometry and material compositions were quantified, and a total uncertainty was derived for each reference critical configuration. The ZPPR-15 experimental records were used to create detailed as-built Monte Carlo models for the reference critical configurations established in ZPPR-15A loading 15, ZPPR-15B loading 88, ZPPR-15C loading 166 and ZPPR-15D loading 185. The as-built models neglect the small, non-uniform gap between matrix halves at full closure, any structures outside the matrix tubes and minor components, e.g., impurities, in some materials. The biases associated with the neglected items were quantified. The experimental value of keff and its associated uncertainty for each reference critical configuration were adjusted to account for biases introduced by items that are neglected in the as-built model for that configuration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗

DOE OSTI2022

Specimen Size and Geometry Effects on the Master Curve Fracture Toughness Measurements of EUROFER97 and F82H Steels

EUROFER97 and F82H are two leading reduced-activation ferritic-martensitic (RAFM) steels for fusion blanket applications. Exposure to the harsh environment of fusion reactors can result in severe degradation of materials fracture toughness (FT). Thus, the post-irradiation evaluation of FT is critical to understanding the material behavior. Due to the space constraint of irradiation facilities, the development of small specimen test techniques (SSTT) is necessary to evaluate the performance of irradiated materials. In this study, we evaluated the specimen size and geometry effects on the ductile-to-brittle transition FT of EUROFER97 batch-3 and F82H-BA12 steels. The specimen thicknesses ranged from 1.65 to 12.7 mm and the geometries included 1.65 mm bend bar, 4 mm mini-compact tension (miniCT), and 0.5T compact tension (CT) specimens. Fracture toughness testing and evaluations were performed using the Master Curve method in the ASTM E1921-19 standard. After size correction to 1T size using the Master Curve method, no specimen size effect was observed between the 4 mm miniCT and 0.5T CT specimens for the Master Curve reference temperature T0Q, while the bend bars yielded a higher T0Q. A strong effect of fatigue precrack front straightness on T0Q for 0.5T CT specimens was observed. The minimum number of specimens needed for each specimen geometry has been determined.

Chen, Xiang↗

NASA NTRS1977

Study of the potential for improving the economics of hydrogen liquefaction through the use of centrifugal compressors and the addition of a heavy water plant

An approach to the liquefaction of hydrogen was developed which permits the application of standard centrifugal compressors in place of reciprocating machines. A second fluid, such as propane, is added to the hydrogen prior to compression to form a mixture having a molecular weight much greater than that of hydrogen alone, so that a standard centrifugal compressor can be used. After compression, the mixture is cooled to cryogenic temperature levels where the propane condenses out of the mixture and is separated as a liquid. Since a small amount of deuterium is produced during hydrogen liquefaction, the potential of recovering deuterium and selling it as a co-product was investigated. Deuterium, in the form of heavy water, can be used in certain nuclear reactors as a neutron moderator to reduce the neutron velocity and enhance the probability of neutron collision with uranium nucleii.

Baker, C. R.↗

NASA NTRS1979

A reactor for more efficient solar cells

Reactor produces highly pure silicon at relatively high temperature of 2,000 K. Process separates liquid silicon product from gaseous coproducts more easily than conventional lower-temperature processes. High production rates may be obtained in relatively small reaction chambers which could include means for collecting or casting silicon ingots.

Fey, M. G.↗

DOE OSTI2024

Modeling and Design of Micro Flux Monitors for Neutron Fluence Measurements

Measuring neutron flux is critical to understanding the characteristics of nuclear reactors and can provide information on the energy spectrum, total neutron fluence, and operational history of a facility. Such measurements are typically performed by irradiating high-purity materials in an operating reactor for a known length of time. By measuring and identifying activation products in the irradiated materials and referencing the energy-dependent cross section of each isotope, the neutron energy spectrum and total fluence can be determined. Historical techniques use a variety of small, individually pressed foils or wires for dosimeter devices, which can result in excessively high levels of radioactivity and complications in handling after irradiation. The present work describes a new neutron dosimetry concept, named the micro flux monitor, that uses semiconductor device fabrication processes to deposit thin films of flux monitoring materials in precise geometries onto a SiC wafer. This approach allows multiple dosimetry materials to be located on a single device, simplifying irradiation and subsequent analysis. The geometry, and therefore mass, of each material can be custom-tailored for a specific application—considering the expected neutron flux, duration of irradiation, and device retrieval time of a facility. Additionally, hundreds of identical dosimetry monitors can be produced simultaneously, reducing fabrication costs. Finally, the extremely thin metal dimensions of these devices could improve the fidelity of neutron fluence measurements by reducing neutron self-shielding and gamma self-attenuation effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Interactive Mesh Generation and Comparison Tools for Nuclear Energy Advanced Modeling & Simulation Phase I. Final Report

The promise of computer simulations for nuclear reactor designs is the opportunity to speed up the design process while also improving safety, providing more detail and higher fidelity, reducing cost, minimizing schedule risk, and avoiding many other potential problems; because simulations have the potential to be much less expensive and time-consuming than traditional development – where scale models and physical tests require construction and long lead times – it is possible to consider more alternatives and complete more comprehensive assessments before the design is finalized and constructed. As more design scenarios are studied by simulation, the speed that simulations promise is only available as long as bottlenecks are addressed for these scenarios. A frequent bottleneck is the amount of human labor required to describe the shape of the objects being designed and decompose the shape into simple elements, such as hexahedra, that are small enough to capture the physical phenomena of interest without being so small that even fast computers cannot perform the simulation in a reasonable amount of time. Recently, techniques for creating all-hexahedral decompositions have been developed, but they require some human input. Many simulation codes prefer all-hexahedral decompositions, so we proposed to evaluate the feasibility of these techniques on nuclear reactor geometries to see whether they required too much human input to be commercially viable. Our study concluded that they can be made viable with some additional software tools to reduce the amount of user input required.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

In-situ coating of silicon-rich films on tokamak plasma-facing components with real-time Si material injection

Abstract Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– 7.5 × 10 19 m −3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s −1 (1– 4.2 × 10 20 Si s −1 ) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO 2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s −1 . The technique is estimated to coat a surface area of at least 0.94 m 2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Volume 10: Flow Test Facilities

Adding additional flow testing facilities has been considered as part of the High Flux Isotope Reactor (HFIR) Sustaining and Enhancing Neutron Science (SENSe) Initiative at Oak Ridge National Laboratory (ORNL) to support HFIR operations and experiments. This prospect has prompted many ideas and discussions regarding potential features, configurations, locations, and applications for the facilities. A working group of ORNL staff members was formed in fiscal year 2022 to recommend one or more configurations to best support future HFIR operations and scientific capacities and to develop order-of magnitude cost estimates and timing. The ideas discussed in this report include options ranging from upgrading existing small-scale testing facilities to building a full-scale HFIR mockup for detailed thermohydraulic testing and fuel assessment.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Performance Demonstration of Self-Powered Neutron Detectors for Steady-State Reactor Operations

The irradiation testing of sensors in reactors is a crucial step towards calibrating and qualifying sensors prior to their deployment in experiments. This report details the process toward qualifying and calibrating custom-designed rhodium-based self-power neutron detectors (Rh-SPNDs) for steady-state reactor irradiations. This process serves to both demonstrate the performance capabilities of Rh-SPNDs as well as to provide experimental data for development of a sensor sensitivity model. Two designs of Rh-SPNDs were tested in various reactors to demonstrate: detection resolution in a low neutron flux environment, a delayed-response compensation technique, output linearity in a large range of neutron flux, and measurement accuracy verified with dosimetry. The detection resolution and compensation technique was demonstrated in the AGN-201m reactor at Idaho State University. The irradiation confirmed the sensors’ capability to perform steady-state operations in a low neutron flux of ~2E8 n/cm 2 -sec. Sensor output linearity coupled with the delayed-response compensation was investigated at the neutron radiography reactor at Idaho National Laboratory. A Rh-SPND was irradiated to neutron fluxes ranging from 2E8 to 2E13 n/cm2-sec range. The measured data demonstrated a wide and linear range of operation with a measured linear sensitivity of 1.0129 ×10 -13 A/W with a correlation-squared value of r 2 =0.9927. The measurement accuracy was investigated at the Advanced Test Reactor Critical reactor. The SPNDs were inserted into a test vehicle with collocated flux wires. Two irradiations with different flux levels were performed, and the SPNDs relative measurement between the two irradiations was calculated to be 1.2613 ± 0.0153 for the small SPND design and 1.1809 ± 0.0108 for the large SPND design. Both SPND measurements fell between the co-axial dosimetry result, which reported 1.218 ± 0.047. Additionally, the preliminary MCNP model for calculating SPND sensitivity was developed in parallel to this work. Modeled neutron spectrum with measured magnitude was used for inputs to determine the simulated SPND output. The results showed an overestimation of signal strength by a factor of 5, which was expected because of model simplification. This leads to future modeling work to account for signal losses from additional physical properties, including high temperature environments for FY-21.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

New High-Strength Ni-based Alloys for High Temperature Service in Liquid Fluoride Salt Environments

The ever-increasing demand for higher system thermal efficiency necessitates the operation of power generation cycles and heat conversion systems for chemical processes at progressively higher temperatures. As the system operating temperature increases, fewer and fewer materials are available with the required mechanical properties and environmental compatibility. This dearth of materials is particularly acute in structural applications at temperatures above 700°C in liquid Fluoride Salt Cooled High-Temperature Reactors (FHRs) and Concentrated Solar Power (CSP) systems where liquid fluoride and/or liquid chloride salts are used for their high thermal capacity and low thermal conductivity to store and transport heat. It has been shown that fluorides and chlorides can degrade the properties of the material used for storage and transport due to corrosive effects thus affecting the performance and lifetime characteristics. Thus, materials used for storage and transport for fluoride/chloride-salt must at the minimum have good compatibility with the salt by exhibiting low corrosion rates at the required temperatures. In addition, since many of the components will be subject to stresses in service, they need to have sufficient strength to resist plastic deformation, and the resistance to time dependent creep deformation at these temperatures, and stress levels. Achieving the combination of these properties required for higher temperatures is particularly challenging considering traditionally, the expected lifetime of some of the components used in these systems are from 30 years (CSP) to 80 years (FHRs). Although 316 stainless steel and associated redox corrosion reduction techniques are being considered for use in the first generational systems, use of 316SS is limited to operating temperatures up to ~ 650°C, thereby resulting in lower reactor efficiencies. Hastelloy ® N, developed at ORNL in the 1950s-60s for molten salt service is currently the leading candidate FHR structural alloy but has inadequate creep properties at high temperatures. Using a computationally-guided approach, Oak Ridge National Laboratory had developed new high strength alloys strengthened by γ’ precipitates (about 2X the strength of Hastelloy ® N) in small laboratory-scale heats that show good resistance to corrosion by fluorides and improved creep rupture life at temperatures up to 850°C (U.S. Patent Application No. 13/833,357 entitled “High Strength Alloys for High Temperature Service in Liquid-Salt Cooled Energy Systems”- granted U. S. Patent No. 9,540,714).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Post-irradiation examination of legacy high burnup fuel to support safety testing

Safety/transient testing to evaluate performance under off-normal conditions is an essential pillar for both the development of Accident Tolerant Fuels (ATF) and the optimization of fuel operation economics beyond current discharge burnups. Among other factors, the successful interpretation of the transient testing results relies upon the knowledge of the initial conditions of the test, including the characteristics of the fuel system under scrutiny. When testing pre-irradiated material, the assumptions that the fuel and the cladding still have the same properties as in the pre-irradiation stage is obviously wrong and could affect the results of the test. This is particularly true the more burnup accumulates in the fuel rod and irradiation progresses. The knowledge of the initial microstructure of both fuel and cladding allows a clearer interpretation of the subsequent transient testing results, provides validation of the physical phenomena underlying the model predictions and eliminates the uncertainties related to the limited knowledge of the sample status before the test. One example is the phenomenon of fine fragmentation that occurs in Light Water reactor (LWR) fuel. During a Loss of Coolant Accident (LOCA) or Reactivity Initiated Transient (RIA) the fuel can severely fragment. During LOCA, high burnup fuel tend to finely fragment, which has raised safety concerns due to the increased likelihood of dispersal of such small particles once the cladding has burst and due to the increased fission gas release. Therefore, efforts have been devoted to the assessment of a pulverization threshold that could determine the conditions under which fine fragmentation is predominant. However, the lack of information regarding the initial conditions of the fuel, and the connections between those conditions and the pre-transient irradiation history, have hindered the development of a fully mechanistic fragmentation and pulverization criterion. The empirical relationships rely on conservative estimations, due to the lack of information on critical material properties and characteristics. More generally, experimental evidence of the irradiation-induced modifications at microstructural scale are necessary to determine the behavior of the material at the macroscopic scale, with the latter being the one of technological interest. Significant progress has been made in the last two decades in the developments of analytical materials science techniques that can be applied to highly radioactive materials, such as high burnup fuels. The availability of new techniques and the improvement of existing ones has enabled investigations previously not possible that can deepen the understanding of the fuel characteristics and properties at high burnup. The better knowledge of material behavior and irradiation-induced phenomena could help the prediction of its performance. In this context, the scope of the present work is to apply a wide portfolio of advanced characterization techniques to determine properties that are relevant for safety and performance. The results are interpreted in the context of engineering scale post-irradiation examinations and available information on the irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Measuring Signatures of Fuel Irradiation in Large Particle Samples

Understanding the operating history of a nuclear facility is one of the key goals in nuclear safeguards, enabling the verification of declared activities. This is accomplished via a number of mechanisms, including the analysis of samples collected at the facility. The best metrics for understanding and verifying the history of nuclear fuel are (1) actinide isotopic content and (2) actinide inter-element ratios. While techniques have long existed for measuring these values with high accuracy and precision in bulk samples, as samples become smaller – such as may be collected in environmental collection – new methods are required both for clean processing of pure samples, and for analysis of these small samples. This study describes a new extraction chromatography technique for recovering U, Pu, and Am from single small aliquots, and advanced mass spectrometry methods for high-precision and high-accuracy analysis of trace and ultra-trace levels of actinides. These techniques are applied to particle samples cut from fuel from a historic nuclear reactor. Ages calculated from Pu–Am chronometry of particles are in excellent agreement with the known operating history of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

LLNL FESP Theory Highlights: August 2024

The 2024 ABOUND SciDAC and BOUT++ combined workshop was held August 5-9 th 2024 at the University of California Livermore Collaboration Center (UCLC) in Livermore. Bringing together leading scientists and researchers from across the globe, this pivotal event focused on advancing plasma physics and boundary plasma dynamics within the context of fusion energy research. Key discussions throughout the meeting highlighted significant advancements in the BOUT++ framework, including enhanced simulations of small Edge Localized Modes (ELMs) and the initiation of integrating the integration of the 5D GEM gyrokinetic turbulence core code with the 2D SOLPS-ITER boundary transport code. These developments are crucial for managing heat loads in fusion reactors and supporting the longevity of plasma-facing components. The event also featured a session on Inter-SciDAC Collaborations, where principal investigators from multiple U.S. FES SciDAC tokamak projects explored opportunities for cross-collaboration. Additionally, the meeting showcased cutting-edge advancements in GPU acceleration and AI/ML technologies, poised to drive the next generation of fusion research. In his closing remarks, Dr. Xueqiao Xu emphasized the importance of the collaborative efforts and discussions that took place, noting their potential to shape future breakthroughs in fusion energy. The event underscored the global nature of the BOUT++ collaboration, with contributions from over 57 institutions worldwide. The 2024 BOUT++ and ABOUND Joint Hybrid Meeting continues to drive forward the research and innovations needed to achieve fusion energy, setting the stage for future collaboration and discovery.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗