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At least 145 records · Page 8

Manufacturing of HEAs at Different Scales

The knowledge gained from melting and processing of high-entropy alloys (HEAs) at different scales is discussed with respect to melt parameters and characteristics. The melting techniques considered are button melting, vacuum induction melting (VIM) and electroslag remelting (ESR). While VIM produces HEAs with enhanced chemical homogeneity, particularly after being subjected to a homogenization heat treatment, and refined grain structure, other concerns arise from elemental contaminants associated with industrial-grade melt stock. Although ESR of the VIM product decreases the concentration of tramp elements, the narrow melt range typically found in HEAs or medium entropy alloys was found to decrease the melt efficiency. The experiments presented were performed on ingots ranging from 100 g to 75 kg.

Detrois, Martin↗

Auxiliary heating and current drive physics for the ST-E1 fusion power plant

This work describes the physics basis for the proposed auxiliary heating and current drive system on the ST-E1 fusion power plant. The ST-E1 flattop plasma considered here is fully non-inductive with a bootstrap fraction of 0.9 and the remaining current driven by EC waves. Using the recently published physics-based optimization method for EC launchers (Lopez et al 2025 Plasma Phys. Control. Fusion 67 055012), we show that the target flattop ECCD can be achieved with a net efficiency of 52 kA MW −1 using fundamental O-mode (O1) with frequency range 160–200 GHz launched from the low-field side top half of the vacuum vessel (LFS top-launch). From considering two candidate rampup scenarios, we conclude that LFS top-launch O1 ECCD can be equally effective during the early stages of plasma operation, although poloidal steering might be needed. X-mode waves injected from the LFS midplane are also shown to be effective for rampup even when T e < 1 keV. We also present modeling results for the pre-conceptual design of an ICRH system proposed for ST-E1. Using TORIC, we find that an ICRH system aiming for 42–48 MHz and toroidal mode number n φ ~ 10 robustly achieves dominant ion damping via Helium-3 minority heating transitioning to second-harmonic Tritium heating. We then show that such waves can be efficiently generated by a 5-strap traveling-wave antenna (TWA) using the Petra-M code. The TWA has a 40–45 MHz passband within which ~60% of the power entering the TWA is coupled to the plasma with the remaining ~40% of the power being transmitted through the TWA and possibly recirculated; the power reflected back into the transmission lines is negligible. This passband structure persists even when the evanescent distance is increased by a factor of two, or when the magnetic-field angle is increased by 30°, demonstrating inherent load resilience that will be crucial for effective ICRH on ST-E1.

electron cyclotron↗

Assessment of Impurities in Neptunium Metal Using Double-Focusing SF-ICP-MS

In the transmutation process, minor actinides such as neptunium, americium and curium obtained from spent nuclear fuel are incorporated into new fuel that can be placed into a fast reactor, where these elements fission into products with shorter half-lives which reduces the health and safety hazards. In order to better understand the all the involved processes, fundamental physical properties of actinides should be investigated, and for that reason, high-purity materials need to be produced. One of the main challenges in characterizing these materials is the lack of pure metal available for study, and lack of analytical methods to determine the purity. Trans-uranic elements are challenging to study because of the need to be handled in specialized facilities, safeguards concerns and they are not easily purchased and/or produced. High-purity Np metal is being produced at the Materials and Fuels Complex (MFC) of Idaho National Laboratory (INL), in a process of direct oxide reduction developed by Squires and Lessing, with a target purity of 99.999%. The goal of this research is to develop methods to determine the impurities in the Np metal, with detection limits low enough to achieve this purity level. In this work, we have used several ICP-MS and ICP OES techniques, but the focus was the method development using a recently acquired Double-Focusing Sector Field Inductively Coupled Plasma Mass Spectrometer (SF-ICP-MS). To produce the Np metal, approximately 20 g of the neptunium oxide powder was calcined at 800 oC for 6 h in a magnesium oxide crucible, which is placed inside a stainless steel crucible, that goes into the Hot Uniaxial Press (HUP) furnace insert. After the calcining step, the Np oxide was mixed with approximately 9 g of calcium metal and 100 g of ultra-dry calcium chloride. The HUP furnace was heated in a ramp rate of 50 oC min -1 up to 1000 oC. When the reaction mixture reached 850 oC, a tantalum stirrer was activated at 100 rpm. After the tantalum stirrer was removed and the mixture cooled overnight, the MgO crucible was opened with a hammer and chisel to retrieve the Np metal button at the bottom of the reaction mixture. Prior to the dissolution process, the surface contaminants of the sample were removed. The Np button was washed in 10 mL of ultra-pure water for 1 min, with a faint reaction observed on the surface, indicating the presence of Ca metal. The water-washed sample was placed in a tube with 2 M HCl for only 3 s, as a vigorous reaction in the button surface was observed. The leachates were saved for further analysis. The sample was then rinsed with ultra-pure water and air-dried prior to dissolution. The cleaned sample was placed in a PTFE beaker and 25 mL of 6 M HCl and 150 µL of 24 M HF were added. After 10 minutes of reaction, the beaker was placed on a hot plate for 1 h and 30 min, followed by the addition of 1 mL of 16 M HNO3 and allowed to react for a further 1 h and 20 min. The sample was removed from heat and 10 mL of HNO3 16 M was added to react with the HCl. The beaker was covered and left for overnight reaction, producing a solution with approximately 20 mL of final volume. A Quadrupole Inductively Coupled Plasma Mass Spectrometer (Q-ICP-MS, Perkin Elmer ELAN DRC II Q-ICP-MS) was used to obtain qualitative mass spectra for the dissolution blank, each of the leachates and the dissolved sample. In the water leachate spectrum mainly Ca and Mg were observed, while the acid wash had a similar peak pattern to the one observed for the Np metal sample, indicating that the process not only removed surface contamination, but also reacted with the sample itself. In the Np material spectrum, Ta and actinides were identified as candidates for major contaminants, as the most intense peaks corresponded to the mass-to-charge ratios of these elements. Minor contaminants included Al, Au, Ca, Cr, Fe, I, Na, Nb, Pb, Sn, Ti, Tl, V, Y. The presence of impurities in the material could have originated from several sources, such as the materials used for the casting process, the use of impure NpO2 or contamination of reagents and materials used for the analysis. Some of the isotopes listed as potential contaminants are susceptible to spectral interferences, so Double-Focusing SF-ICP-MS (Nu Instruments Attom ES HR-ICP-MS) was used for the quantitative analysis. For each isotope determined, several items were evaluated: raw signal of a blank solution, sensitivity and determination coefficient of calibration curves, possibility of interferences and comparison of results obtained for at least two resolving powers (RP): 300 and 4000. The evaluation of the signal in a blank solution can indicate if a spectral interference from the plasma gas and/or the solvent was successfully resolved (commonly observed for 52Cr+/40Ar12C+, for example). It is important to highlight that a signal reduction is expected when a higher RP is used, as the ion beam transmission is decreased by the use of narrower slits

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Tool Chain for Simulations of Bi-Filar Coil Winding for Fast Quench Protection

The advancement of high-field magnets utilizing high-temperature superconductors (HTS) brings about complex challenges, especially in quench detection and protection. Traditional methods often fall short due to the inherently slow quench propagation in HTS materials. One promising approach to overcome this involves using a bifilar winding configuration, where two conductors are placed side by side. Under normal operation, they function in series, but during a quench event, they switch to an anti-parallel mode. This shift reduces the differential inductance of the coil to near zero, enabling rapid current oscillations through a capacitor discharge. The resulting high-frequency current flow leads to swift, uniform heating, triggering a full-coil quench within microseconds. Moreover, the strong mutual coupling between the two windings significantly reduces electrical noise in voltage measurements. In this work, we explore the viability of this concept by designing, constructing, and testing a REBCO bifilar racetrack coil in liquid nitrogen. We also present a validated simulation model that closely mirrors the coil's dynamic behavior under these conditions, aligning well with experimental observations.

Jayathilaka, Rehan [Northern Illinois U.]↗

Superconducting Magnet Qualification Methodology for the Material Plasma Exposure Experiment

The Material Plasma Exposure eXperiment (MPEX), currently under design, is a new linear plasma device to advance the understanding of plasma-material interactions through the generation and delivery of plasmas as they are expected in future fusion reactor divertors. MPEX will be a steady-state device to study high-fluence exposures of plasma-facing materials and components. The requirements for the magnetic field at the target and the heating stages make the application of superconducting coils necessary. Conceptual designs for the superconducting magnets have been developed, and multiple cryostats with warm bore diameters of either 65 cm or 156 cm are envisioned to facilitate their integrated and timely assembly with other systems such as vacuum, water cooling, and RF power. Although design, fabrication, and testing for the magnets as stand-alone units are straightforward, challenges will arise during the integration of the system. Two different field profiles will be used during operation. The magnetic field where the electron cyclotron heating occurs needs to operate at both 1.25 and 2.5 T. It is critical that the magnets all share the same magnetic axis and alignment. The mutual inductance between cryostats will affect the quench behavior of the system. Also, cryostat-to-cryostat forces can be as large as 700 kN, and the magnitude and direction will change depending on which coils are energized. The design of the system must take those characteristics into account along with the quench scenarios. This paper describes the qualification approach that will be used to determine whether stand-alone tests can be used to ensure the success of the integrated system. Fiducials will be used to define the location of the magnetic axis for each cryostat to ensure proper alignment. Quench tests of a single magnet will be performed at a current above the normal operating current to account for additional stored energy from the mutual inductance to adjacent cryostats. Also, a 1018 steel plate will be mounted on either end of a cryostat to simulate the cryostat-to-cryostat forces. Requirements for the size and location of the steel plates are described.

Burkhardt, Earle↗

Transformer with integral inductor

In accordance with one embodiment is a transformer with a core comprising a perimeter portion and central intervening portion. The central intervening portion is separated from the perimeter portion by air gaps, creating an opening on either side of the intervening portion. A primary winding and secondary winding are wound around the central intervening portion of the core. The primary winding is capable of electromagnetic interaction with the secondary winding. A pair of ferrite members arranged outward from a central axis of the central intervening portion of the core and increases a series inductance with the primary winding. In accordance with another aspect of the disclosure, each ferrite member may have an air gap associated with the core to facilitate heat dissipation from the transformer.

Rastogi, Sagar K.↗

Report on the Analysis of Plutonium-238 Oxide, Campaign #P5PO

Oak Ridge National Laboratory (ORNL) is currently producing Plutonium-238 oxide for NASA space programs. Through the irradiation of NpO 2 pellets in the High Flux Isotope Reactor (HFIR), the production of Pu-238 oxide will be used for electrical power and heat on NASA spacecraft such as the Curiosity Mars Rover and the Voyager 1 and 2. The Transuranium Analytical Laboratory (TAL) leads the analytical effort for this project, which includes the dissolution of plutonium product oxide (PUP) and the analysis of the dissolved product solution using both radiological and inductively coupled plasma mass spectrometry (ICP-MS) techniques that are required to meet product specifications. The Transuranic Analytical Laboratory received three samples of Pu-238 oxide for analysis in January of 2022. This report documents the analytical protocols performed by the TAL along with the results produced from said analytical protocols. This report also includes a comparison with the general-purpose heat source (GPHS) specifications.

07 ISOTOPE AND RADIATION SOURCES↗

Active Lithium Injection for a Real Time Control of the Divertor Heat Flux for Fusion Devices

When the local heat flux exceeds specified flux limit, tungsten PFC surfaces can be damaged, which is not acceptable for a reliable reactor operations. The divertor PFCs are typically designed for a specific heat flux limit usually assuming an average steady-state heat flux which is typically 5–10 MW/m 2 . However, in addition to steady-state heat flux, fusion reactor divertor PFCs could experience transient heat fluxes such as ELMs and/or other magnetic reconnection events which can deposit large transient heat fluxes onto the divertor PFCs. The transient divertor heat flux could be significantly larger than the steady-state heat flux which could damage the solid PFC surfaces. The divertor heat flux can be subjected to additional complications such as the uncertainties in the the divertor strike point heat flux projection. Moreover, there are additional experimental observations of non-axisymmetric power flux which can occur under non-axisymmetric magnetic perturbations. The liquid lithium (LL) PFCs is more resilient against such transient heat fluxes as they could evaporate LL as needed and the lost LL can be then replenished afterward. In this paper, we analyze a case for a transient divertor heat pulse of 1 MJ in 10 ms for a ITER-size reactor. This is a small perturbation (~ 0.1%) to the expected plasma stored energy compared to the previously analyzed case of 20 MJ heat pulse. Even with this relatively modest heat pulse, the LL surface undergoes ~ 100 °C temperature rise. However, the resulting LL surface heating without rapid cooldown mechanism could lead to excessive LL evaporation continuing well after the transient heat flux resulting in a significant Li injection of ~ 0.6 mol in about a 200 ms period. This amount of Li injection could cause plasma dilution and performance degradation. On the other hand, an active Li injection capability if optimized could prevent the LL surface temperature rise and thus reducing subsequent Li evaporation into the plasma by a factor of 7 compared to the passive LL PFC case. A crucial tool of active Li injection is a rapid response pellet injector which could inject light impurity pellets before the excessive heat flux could reach the divertor plate causing serious damage. A simple pellet ablation model suggests a favorable pellet deposition profile for smaller ~ 0.1 mm radius pellet with ~ 10–20 m/s velocity. Moreover, if it is possible to inject from the private flux region, the pellet injection efficiency into the high heat flux strike point region can be as high as 80% compared to ~ 50% for the injection from outer radius region. The pellet deposition efficiency can be further improved by designing a shell-pellet which can burst when a certain ablation fraction is reached. Finally, a possible implementation technique using an inductive pellet injector with a rapid time response of a few msec is proposed here which can be tested in NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Length and Seed Current Scaling of a Mark X HG driven, 8 inch PBX9501 RancheroS FCG system: II

This report is intended to examine the ultimate operating limits of the combined Pt. 88 capacitor bank (CB) + MarkX Helical Generator (HG) driving a RancheroS Flux Compression Generator (FCG) with various physics experiment load inductances. As such, many degradation mechanisms are neglected in this study. In actual use, degradation of the HG output current delivered to the FCG will probably occur. Internal changes in the HG can occur due to both Ohmic heating in the generator leading to a time varying series resistance [R(t)] such as that seen in the first trial (Shot 0) of the MarkX, as well as any misbehavior, such HG internal turn to turn shorts such as seem to have occurred in subsequent MarkX tests. Both reduce the output current from the HG that is used to seed the FCG. The results discussed here are intended to probe the limits of the RancheroS driving various loads, in an ideal situation in which none of the HG or system degradation mechanisms prior to the FCG armature first motion (FM) reduce the seed current. The HG is modeled simply as a time varying inductance [L(t), as measured in Shot 0], with a constant internal resistance [R(t) = 0.2mOhm]. The RancheroS itself is fully modeled, including all FCG and load degradation mechanisms, thus probing the ultimate capabilities of the full system, independent of problems upstream of the FCG.

47 OTHER INSTRUMENTATION↗

Characterization of Self-Magnetic Pinch (SMP) radiographic diode performance on RITS-6 at Sandia National Laboratories: 1) Diode Dynamics, DC Heating to extend Radiation Pulse

Radiographic diodes focus an intense electron beam to a small spot size to minimize the source area of energetic photons for radiographic interrogation. The self-magnetic pinch (SMP) diode has been developed as such a source and operated as a load for the RITS-6 Inductive Voltage Adder (IVA) driver. While experiments support the generally accepted conclusion that a 1:1 aspect diode (cathode diameter equals anode-cathode gap) delivers optimum SMP performance, such experiments also show that reducing the cathode diameter, while reducing spot size, also results in reduced radiation dose, by as much as 50%, and degraded shot reproducibility. Analyzation of the effective electron impingement angle on the anode converter with time made possible by a newly developed dose-rate array diagnostic indicates that fast-developing oscillations of the angle are correlated with early termination of the radiation pulse on many of the smaller-diameter SMP shots. This behavior as a function of relative cathode size persists through experiments with output voltages and currents up to 11.5 MV and 225 kA, respectively, and with spot sizes below ~ few mm. Since simulations to date have not predicted such oscillatory behavior, considerable discussion of the angle-behavior of SMP shots is made to lend credence to the inference. There is clear anecdotal evidence that DC heating of the SMP diode region leads to stabilization of this oscillatory behavior. This is the first of two papers on the performance of the SMP diode on the RITS-6 accelerator.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Infrared laser ablation of poly(vinylidene fluoride): The Loss of HF

The mechanism and kinetics of infrared laser ablation of poly(vinylidene fluoride) (PVDF) in vacuum have been investigated to understand how this can be used as a polymer processing technology. The laser heats the surface and initiates decomposition of macromolecules in the molten polymer layer which rapidly leads to production of an ablative flow of polymer decomposition products. There is only a short induction period of up to 2 s before ablation begins. After the initial period, the rate of mass loss increases linearly with the time of laser exposure. Spherical particles with an average diameter of about 100 nm are formed on the surface of the powder coating obtained by laser ablation. Here, the particle distribution ranges from 50 nm to 300 nm. The main gaseous product of ablation is HF from dehydrofluorination. A second important ablation process is the formation of gaseous carbenes generated during secondary reactions. Proposed ablation reactions due to laser-induced thermal dehydrofluorination occur in two stages. In the first stage, a double bond appears in PVDF due to direct loss of HF. Subsequent detachment of HF leads to the appearance of both triple bonds and cumulene bonds.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Parameter space for magnetization effects in high-energy-density plasmas

Magnetic fields are well known to affect the evolution of fluids via the J × B force, where J is the current density and B is the magnetic field. This force leads to the influence of magnetic fields on hydrodynamics (magnetohydrodynamics). Magnetic fields are often neglected in modeling of high-energy-density plasmas, since J × B is very small compared with the plasma pressure gradients. However, many experiments lie in a separate part of parameter space where the plasma is indirectly affected via magnetization of the heat flux and charged particle transport. This is true even for initially unmagnetized plasmas, since misaligned density and temperature gradients can self-generate magnetic fields. By comparing terms in the induction equation, we go on to estimate the regions of parameter space where these self-generated fields are strong enough to affect the hydrodynamics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Power Electronics Manufacturing Improvements for Heavy-Duty Fuel Cell Vehicles

The Marel Power Solutions project, funded by the U.S. Department of Energy under Award DE-SC0023801, focused on advancing manufacturing techniques for power electronics in heavy-duty fuel cell vehicles. The research aimed to enhance system efficiency, reduce costs, and support broader adoption of hydrogen fuel cell technology. Key areas of investigation included power topology, thermal modeling, system architecture, and accessibility through software tools. Key Accomplishments: 1. Power Topology: - Developed an interleaved boost converter with optimized phase count, leveraging Marel’s proprietary Power Stacks. - Achieved reduced parasitic inductance and resistance, enabling high efficiency in DC-DC converters. 2. Thermal Modeling: - Integrated innovative cooling systems into compact Silicon Carbide (SiC) modules. - Simulations demonstrated the ability to dissipate significant heat (up to 7.5 kW), ensuring device reliability under heavy loads. 3. System Architecture: - Utilized simulation tools to analyze the impact of various fuel cell and vehicle parameters on efficiency. - Highlighted the role of smaller, modular improvements, such as enhanced DC-DC converters, in achieving system-wide gains. 4. Accessibility: - Evaluated and implemented MATLAB/Simulink code generation tools for real-world hardware applications. - Demonstrated the potential for rapid prototyping of custom power systems with reduced development costs. Impact and Benefits: - Efficiency and Cost Reduction: Marel’s cooling technology enhances SiC die performance, reducing the number of dies required and overall system size. - Scalability and Flexibility: The innovations support tailored solutions for diverse applications, from mass transit to mining vehicles. - Sustainability: The research promotes the integration of electrification technologies, helping meet rising energy demands sustainably. Conclusion: The project’s outcomes advance the state of power electronics for hydrogen fuel cell vehicles, enabling more efficient, compact, and cost-effective solutions. These developments lay a foundation for future innovation, contributing to the broader adoption of clean energy technologies in transportation and other industries.

08 HYDROGEN↗

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Characterization of self-magnetic pinch (SMP) radiographic diode performance on RITS-6 at Sandia National Laboratories. I. Diode dynamics, DC heating to extend radiation pulse

Radiographic diodes focus on an intense electron beam to a small spot size to minimize the source area of energetic photons for radiographic interrogation. The self-magnetic pinch (SMP) diode has been developed as such a source and operated as a load for the six-cavity radiographic integrated test stand (RITS-6) inductive voltage adder driver. While experiments support the generally accepted conclusion that a 1:1 aspect diode (cathode diameter equals anode–cathode gap) delivers optimum SMP performance, such experiments also show that reducing the cathode diameter, while reducing spot size, also results in reduced radiation dose, by as much as 50%, and degraded shot reproducibility. Analysis of the effective electron impingement angle on the anode converter with time made possible by a newly developed dose-rate array diagnostic indicates that fast-developing oscillations of the angle are correlated with early termination of the radiation pulse on many of the smaller-diameter SMP shots. This behavior as a function of relative cathode size persists through experiments with output voltages and currents up to 11.5 MV and 225 kA, respectively, and with spot sizes below approximately few millimeters. Since simulations to date have not predicted such oscillatory behavior, considerable discussion of the angle behavior of SMP shots is made to lend credence to the inference. We report there is clear anecdotal evidence that DC heating of the SMP diode region leads to stabilization of this oscillatory behavior. This is the first of two papers on the performance of the SMP diode on the RITS-6 accelerator.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Time dependent model for non-inductive ECH X-I current ramp-up for SHPD tokamak facility

Non-inductive (NI) plasma current start-up and ramp-up is an important research topic for spherical tokamak (ST) based reactors and fusion pilot plant (FPP). For a compact FPP, the OH flux availability is highly restricted due to its compact geometry. Efficient fundamental extraordinary mode (X-I) electron cyclotron heating (ECH) current start-up and ramp-up regime was identified for a reactor-like high toroidal magnetic field range which has more than a hundred times higher current drive efficiency compared to more conventional ECH methods for the relevant start-up temperature range. High current drive efficiency is possible due to the strong X-I fundamental ECH interaction only with unidirectional passing electrons constrained by the wave accessibility conditions. Here, we extend the X-I electron cyclotron current drive (ECCD) investigation to a time dependent model to simulate the non-inductive current ramp-up to 10 MA for the Sustained High-Power Density (SHPD) facility. As the X-I ECCD driven current I EC rises, due to the back EMF driven negative current, the net plasma current Ip rises more slowly with the current resistive time scale. For tokamak confinement time (both L-mode and H-Mode) which tends to rise with I p , a positive feedback results and even with constant applied ECH power, T e0 , I EC , and I p can continue to rise to very high values. However, in a realistic situation, the T e0 rise should saturate due to a number of factors such as enhanced core radiation and increased power loss at high temperature. To simulate this effect, we adopt a maximum T e0 model which would limit the temperature rise to certain T e0 . With this model, we investigated the current ramp-up for various ECH power levels and the maximum T e0 of 15, 20, and 25 keV. We find that while the power required is reduced with increasing T e0 limit due to increased current drive efficiency, the time to reach 10 MA tends to go up due to the reduced plasma resistivity for the higher T e0 limit. We also find for a given Te0 limit, the time to reach 10 MA tends to be reduced by increasing the applied ECH power by over driving the current ramp-up where I EC is driven at significantly higher level than 10 MA. While the current ramp-up time may not be an issue for the steady-state reactor systems, if it is desirable to minimize the current ramp-up time, it is prudent to have a sufficient ECH power for current over-drive and have some T e0 limiting tools such as impurity seeding for enhanced radiation. In conclusion, a well-controlled NI ECH start-up also has a potential of improving the tokamak start-up reliability and avoid run-away electrons while the NI off-axis current drive could enhance MHD stability and plasma performance improvements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Effects of non-thermal termolecular reactions on wedge-induced oblique detonation waves

The shock-induced combustion ramjet (Shcramjet) based on oblique detonation waves (ODWs) is among the promising choices for hypersonic propulsion systems. An understanding of the ignition, propagation, and stability of ODWs is critical to harnessing their propulsive potential. In such high speed reacting flows, there is a high probability of occurrence of non-thermal reactions due to the presence of non-trivial amounts of highly reactive radicals including H, O and OH apart from O 2 as demonstrated recently [M. P. Burke, S. J. Klippenstein, Nat. Chem. 9 (2017) 1078–1082, Y. Tao, A. W. Jasper, Y. Georgievskii, S. J. Klippenstein, R. Sivaramakrishnan, Proc. Combust. Inst. 38 (2021) 515–522]. The present work focuses on examining the initiation, propagation and structure of oblique detonation waves in stoichiometric H 2 -air mixtures through numerical simulations with and without non-thermal reactivity on a two-dimensional adaptive grid. Non-thermal reactions were included in the macroscopic kinetic model as chemically termolecular reactions facilitated by the H + OH radical-radical recombination and the H + O 2 radical-molecule association reactions. Since, the non-thermal reactions result in a corresponding decrease in the reaction fluxes of the incipient recombination/association reactions, an additional simulation was performed by applying corrections to the respective incipient recombination/association rate constants using the methodology demonstrated by Tao et al. [Y. Tao, A. W. Jasper, Y. Georgievskii, S. J. Klippenstein, R. Sivaramakrishnan, Proc. Combust. Inst. 38 (2021) 515–522]. Results show that, under ODWE relevant conditions, non-thermal reactivity fundamentally alters the induction length, intensity as well as the structure of the ODW. Specifically, it is found that non-thermal reactivity leads to a noticeable reduction in initiation length and a simultaneous increase in instantaneous peak heat release rate and the degree of unsteadiness of the ODW. Finally, statistical analysis of key thermo-chemical variables is performed to elucidate the important species as well as reactions responsible for the observed variations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Physics basis for the reference flat-top plasma scenario in the ST–E1 fusion power plant

As part of the U.S. Department of Energy’s Milestone-Based Fusion Energy Development Program, Tokamak Energy has completed the pre-concept design of the ST–E1 fusion power plant. ST–E1 is envisaged to operate in two phases: a pilot plant phase, targeting sustained net power production of 300 - 500 MWe for a duration >1 hr, followed by a commercial power plant phase targeting steady-state operations and a normalised overnight capital cost of ⩽12 000 $\$$/kWe. The design process adopted was highly iterative, integrating all major plant systems and progressing in a phased fidelity approach. At the pre-conceptual stage, the emphasis has been on exploring the design space, identifying the main system-level trade-offs, and making the key decisions that define the overall plant concept, rather than optimising a single operating point. This paper, part of a focused collection detailing the ST–E1 pre-concept design, addresses the development of a series of reference flat-top plasma operating points for the pilot plant phase. A modelling workflow was established to develop and assess candidate plasma design points and explore key dependencies. The workflow includes integrated core plasma modelling, magnetohydrodynamic (MHD) stability assessment, equilibrium generation, scrape-off-layer and exhaust modelling, heating & current drive design and optimisation, and turbulent transport modelling. Using this framework, the impact of several key parameters on the flat-top operating space was investigated, including the density limit, core radiation fraction and divertor power loading, level of external heating and curent drive power and assumed pedestal characteristics. The MHD stability, controllability and micro-stability characteristics of these plasmas were also analysed. These investigations informed the definition of a set of fully non-inductive, flat-top reference operating points that satisfy the high-level ST–E1 mission, including a low and high density case, a case that is stable to resistive wall modes and a case with reduced divertor power loading.

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