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One technique for mitigating disruptions in a tokamak is shattered pellet injection (SPI). SPI is a process in which a large solid pellet consisting of deuterium, neon, or argon is desublimated in ...

Shattered pellet injection (SPI) has been chosen as the baseline disruption mitigation system on ITER due to its ability to rapidly inject material deep into the plasma to greatly increase the plas...

Reliable mitigation is necessary to eliminate the detrimental effects of a disruption event in large high-current tokamaks such as ITER. To avoid serious damage to plasma-facing components during t...

Mitigating disruptions is essential for the longevity of future large tokamak experimental devices and reactors. Currently, shattered pellet injection (SPI) technique is the most effective mitigation technique found thus far, and has been chosen for the baseline disruption mitigation (DM) system for ITER. To optimally design SPI systems, the survivability of the pellet throughout the pre-shatter flight and the resulting shatter spray must be better understood. Experimental tests of low-angle single strike impacts of neon and argon pellets were conducted to determine the minimum normal kinetic impact energy that pellets can withstand throughout guide tube travel. Knowing the maximum normal kinetic energy that pellets of relevant materials and temperatures can withstand during flight will allow for an optimal SPI system design. Characterization of the downstream shatter spray was performed for various shatter tube geometries using pure argon and neon pellets. The experimentally measured post-shatter fragment size distribution is compared to theoretical models. The most accurate model found from this comparison is a statistics-based model for brittle material shattering that is correlated with the relevant physical parameters of SPI pellets. Extrapolation of the model to ITER size pellets is presented.

A continuous pellet fueling system (CPFS) is currently being designed at the Oak Ridge National Laboratory (ORNL) for the long pulse operation of the Wendelstein 7-X (W7-X) stellarator. The purpose of the CPFS is to provide deep continuous fueling for feedback-controlled high-density operation and mitigation of predicted hollow density profiles. The system will provide the capability to inject cylindrical pellets of solid hydrogen or deuterium into the plasma core, with flexibility to vary the pellet size, velocity, and injection frequency. Pellets are nominally of 3 mm in diameter and have a length between 1 and 4 mm. The heart of the CPFS is a vertically oriented, twin-screw extruder, cooled by three Gifford-McMahon cryocoolers in parallel, designed to form a continuous filament of hydrogen or deuterium. The filament width, which determines the pellet length, can be adjusted by means of a variable nozzle driven by a linear actuator at the base of the extruder. Coupled to the nozzle is a solenoid-operated gas gun and cutter assembly. A pneumatic propellant valve pulses a burst of ~60 bar helium to accelerate the cut pellet into W7-X. Three gaps in the guide tubes provide pumping locations to remove the helium propellant before it reaches the plasma. The maximum velocity of the pellet is limited by its ability to survive navigating the guide tube trajectory intact. A microwave cavity located within the guide tube provides the capability to measure the pellet size and velocity. The mechanical and thermal designs of the W7-X extruder, adjustable nozzle, and gun and cutter assembly design are described. A guide tube design and experimental pellet survivability test results are presented.

A digital holography device is currently undergoing development at the Oak Ridge National Laboratory for the purpose of measuring surface topography, with the goal of deployment as a real-time plasma-facing component diagnostic for the study of materials that could be utilized in a nuclear fusion device. The holography system utilizes one or two lasers depending on the scale of surface features under measurement. Measurements of surface roughness were performed in a single-laser mode and compared with the data from profilometry, with a linear correlation of increased holographic measurement fidelity as surfaces became smoother. Characterization of the dual-laser operating mode has been performed via surface measurement of stainless steel targets with “stair-step” features in various sizes. Results demonstrated that surface features with known sizes as small as $25.4~\mu \text{m}$ could be resolved. Measurements were within $\sim 55~\mu \text{m}$ or less deviation from the actual sizes, and measurement accuracy was improved as feature size was increased, corresponding to the effect of noise becoming less pronounced. A target exposed to plasma generated by an electrothermal (ET) arc source was analyzed with flat-field correction and averaging of sequential image frames to demonstrate the improved measurement quality in preparation for future use of holography on ET arc-exposed targets.

A series of experiments have been executed at JET to assess the efficacy of the newly installed shattered pellet injection (SPI) system in mitigating the effects of disruptions. Issues, important for the ITER disruption mitigation system, such as thermal load mitigation, avoidance of runaway electron (RE) formation, radiation asymmetries during thermal quench mitigation, electromagnetic load control and RE energy dissipation have been addressed over a large parameter range. The efficiency of the mitigation has been examined for the various SPI injection strategies. The paper summarises the results from these JET SPI experiments and discusses their implications for the ITER disruption mitigation scheme.

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step. © 2019 Elsevier B.V.

Erosion characteristics of tungsten-alternative plasma-facing materials (PFMs) were tested under high heat flux conditions in the electrothermal plasma source facility at Oak Ridge National Laborat...

Mitigation of disruption events in future high energy density tokamaks is essential for machine longevity. The creation of runaway electrons, large electromagnetic forces, and high localized heat l...

To fulfill the increasing needs of diagnostic support for researchers in plasma technology, a portable diagnostic package (PDP) equipped for both laser Thomson scattering (TS) and optical emission spectroscopy has been designed and constructed at Oak Ridge National Laboratory (ORNL), aiming to measure the temperature and number density of electrons and temperatures of ions in plasma devices. The PDP has been initially implemented on a high density and low temperature electrothermal arc source (ET-arc) at ORNL to test its TS capability. TS from the plasmas in the ET-arc has been obtained using the PDP. The electron temperature and number density were determined from TS spectra. These results were then compared to measurements from previous studies on the ET-arc. The TS diagnostic measured 0.8 ± 0.1, 1.3 ± 0.2, and 0.7 ± 0.1 eV and (4.4 ± 0.5) × 1021, (5.9 ± 0.7) × 1021, and (4.3 ± 0.5) x 1021 m-3, respectively, from three lines of sight that transect the plasma column.

A diagnostic system, which has a design goal of high-portability, has been designed at Oak Ridge National Laboratory (ORNL). This project aims at providing measurements of key plasma parameters (ne, Te, ni, Ti) for fusion-relevant devices, utilizing Thomson scattering (TS) and optical emission spectroscopy (OES). The innovative design employs mostly commercial off-the-shelf instrumentation and a traveling team of researchers to conduct measurements at various magnetic-confinement plasma devices. The TS diagnostic uses a Quantel Q-smart 1500 Nd:YAG laser with a 2ω harmonic generator to produce up to 850 mJ of 532 nm laser pulses at 10 Hz. Collection optics placed at the detection port consists of an 11 × 3 optical fiber bundle, where the TS diagnostic uses an 11 × 1 subset array of the fibers, the OES diagnostic uses another 11 fibers, and the remaining fibers are available to the host institution. The detection system is comprised of two separate IsoPlane-320 spectrometers with triple-grating turrets of various line spacing and two PI-MAX 4 intensified CCD detectors, used simultaneously to measure a broad range of ion, impurity, and electron parameters. The self-contained diagnostic package also includes a data processing and storage system. The design and initial implementation of the TS-OES diagnostic system are described. The experiments from the proof-of-principle operation of the portable package on a high density (∼2.5 × 1022 m−3) and low-temperature (∼5 eV) electrothermal arc source at ORNL are also discussed.

Digital holography has been proposed to fulfill a need for an imaging diagnostic capable of in situ monitoring of surface erosion caused by plasma–material interaction in nuclear fusion devices. A digital holography diagnostic for 3D surface erosion measurement has been developed at Oak Ridge National Laboratory with the goal of deployment on a plasma device. A proof-of-concept in situ demonstration is planned which would involve measurement of plasma erosion on targets exposed to an electrothermal arc source. This work presents the results of an ex situ characterization of the capability and limitations of holographic imaging of targets exposed to the arc source. Targets were designed to provide a fiducial for comparison of deformed and unaffected areas. The results indicated that the average net erosion was ∼150 nm/plasma exposure, which is expected to be within the diagnostic’s measurement capacity. Surface roughness averages determined by holographic image analysis showed good agreement with measurements taken with a profilometer. The limit of the holography diagnostic’s x–y spatial resolution was characterized by comparison with scanning electron microscope imaging.

Extensive investigations, within the EUROfusion Work Package "Tritium, Fuelling and Vacuum", indicate that sufficiently deep fuel deposition inside H-mode plasmas of the EU-DEMO tokamak requires injection of fuel pellets from the High Field Side (HFS) at speeds ≳ 1 km/s. To implement this, two different approaches are being pursued: one makes use of “conventional” curved guide tubes, featuring large bend radii (≳ 6 m), to transport 1 km/s pellets to the HFS while trying to preserve their mass and integrity; the other explores the feasibility of injecting high-speed (≳ 2 km/s) pellets from the HFS, along "Direct-Line-of-Sight" (DLS) paths. This paper focuses on the latter approach. Recent tests with an existing ENEA-ORNL high-speed injector have confirmed that the trajectories of free-flight pellets, travelling under vacuum at speeds up to 2.4 km/s, spread within an angle ≲ 0.68°. Despite their small scatter cone, free-flight pellets may require too much cut off volume of the Breeding Blanket (BB), due to the large distance between the injector and the plasma. The introduction of a straight DLS guiding tube transporting the high-speed pellets, to avoid significant loss of BB material, has been investigated. The existing ENEA-ORNL injector has been modified to accommodate a 10 mm i.d. DLS guide tube, and intact pellets have been consistently delivered downstream of the guide at speeds up to 2.6 km/s, with remarkably reduced scatter cone, thus showing the viability of this innovative approach.

Cryogenic pellets are used for injection into fusion plasmas to add fuel to build up density and replace the ions lost from fusion reactions and imperfect confinement in the plasma. These pellets a...

Future large tokamaks, such as ITER, will require a reliable technique for rapid energy dissipation to mitigate harmful effects from disruptions. Two main methods developed for disruption mitigation are massive gas injection and shattered pellet injection (SPI). Argon and neon are favorable materials for both injection methods. When launching pellets with SPI, it has proven difficult to launch intact pellets of pure argon and/or neon owing to their high material strength at cryogenic temperatures. In this work, we compare two methods of launching relatively high-Z pellets. An electrothermal plasma source is an experimental alternative to the fast opening, high-pressure, gas valve. The electrothermal source was used to launch Lexan™ pellets with approximately the same size and mass of comparable mixed gas (D2 and Ne) cryogenic pellets launched by gas guns. We describe comparisons of achieved pellet velocities, energy efficiencies of each system, and the implications of implementing each respective ...

There are currently few viable diagnostic techniques for in situ measurement of plasma facing component erosion. Digital holography is intended to fill this gap. Progress on the development of single and dual CO2 laser digital holography diagnostics for in situ plasma facing component erosion is discussed. The dual laser mode’s synthetic wavelength allows the measurable range to be expanded by a factor of ∼400 compared to single laser digital holography. This allows the diagnostic to measure surface height changes of up to 4.5 μm in single laser mode and up to 2 mm in dual laser mode. Results include ex situ measurements of plasma eroded targets and also dynamic measurements of nm and μm scale motion of a target mounted on a precision translation stage. Dynamic measurements have successfully been made with the system operating in both single and dual laser modes, from ∼50 nm to ∼4 μm in single laser mode and up to ∼400 μm in dual laser mode (limited only by the stage speed and camera acquisition duration). These results demonstrate the feasibility of using digital holography to characterize plasma facing component erosion dynamically, i.e., during plasma exposure. Results of proof-of-principle in situ digital holographic measurements of targets exposed to an electrothermal arc plasma source are presented.

The technology for producing, accelerating, and shattering large pellets (before injection into plasmas) for disruption mitigation has been under development at the Oak Ridge National Laboratory for several years, including a system on DIII-D that has been used to provide some significant experimental results. The original proof-of-principle testing was carried out using a pipe gun injector cooled by a cryogenic refrigerator (temperatures $\sim 8$ –20 K) and equipped with a stainless steel tube to produce 16.5-mm pellets composed of either pure D2, pure Ne, or a dual layer with a thin outer shell of D2 and core of Ne. Recently, significant progress has been made in the laboratory using that same pipe gun and a new injector that is an ITER test apparatus cooled with liquid helium. The new injector operates at $\sim 5$ –8 K, which is similar to temperatures expected with cooling provided by the flow of supercritical helium on ITER. An alternative technique for producing/solidifying large pellets directly from a premixed gas has now been successfully tested in the laboratory. Also, two additional pellet sizes have been tested recently (nominal 24.4 and 34.0 mm diameters). With larger pellets, the number of injectors required for ITER disruption mitigation can be reduced, resulting in less cost and a smaller footprint for the hardware. An attractive option is longer pellets, and 24.4-mm pellets with a length/diameter ratio of $\sim 3$ have been successfully tested. Since pellet speed is the key parameter in determining the response time of a shattered pellet system to a plasma disruption event, recent tests have concentrated on documenting the speeds with different hardware configurations and operating parameters; speeds of $\sim 100$ –800 m/s have been recorded. The data and results from laboratory testing are presented and discussed, and a simple model for the pellet solidification process is described.

Shattered pellet injection is of the most attractive way to mitigate the plasma disruption in fusion research facilities up to now. DIII-D has already utilized it and achieved very positive results. ITER has decided to adopt this technology for the DMS (disruption mitigation system) for PFPO-1 (Pre-Fusion Power Operation phase 1). The validation between simulation code and experiment, and continuous engineering development need to be carried out to meet the DMS’s requirement and reliability. KSTAR (Korea Superconducting Tokamak Advanced Research) is a possible candidate to test the urgent issues of plasma disruption for ITER. KSTAR can install two injectors in toroidal opposite positions. For this work, ORNL (Oak Ridge National Laboratory) will provide the two injectors, the shatter tubes and auxiliary systems. NFRI (National Fusion Research Institute) is preparing the infrastructure of a pumping system, control and data acquisition system, and arranging the location of diagnostic and the heating systems. This presentation describes the basic requirements and the engineering challenges to be solved for successful deployment and operation of multiple SPI injectors in 2019.

Future long-pulse magnetic confinement fusion reactors will require density and isotopic mixture control using steady-state repeating pellet injectors. For high-energy density burning plasmas, pell...

The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed in LabVIEW. More recently, a remote-control system has been set up, based on RealVNC®, which allows to operate the IPI from a control room in Italy. Tools for data transfer and storage into ENEA ICT area have also been provided. A Staging, Storage and Sharing system, named E3S, developed using OwnCloud as architectural component, is used for file synchronization and sharing of the data acquired by the diagnostic systems. It provides a homogeneous platform able to store and share heterogeneous data produced by many data acquisition systems in large nuclear fusion experiments. This paper reports about the implementation of the IPI remote control, and presents the application of E3S to this specific case, allowing easy storage and sharing of experimental data onto a wide-area distributed file-system, as well as remote data access via web-services based on MDS+ tool, integrated with MySQL metadata. A performance analysis of the architectural components is also introduced.

The technology to form and shoot high-Z cryogenic solid pellets mixed with deuterium using a gas gun that are shattered upon injection into a plasma has been developed at Oak Ridge National Laboratory for mitigating disruptions. This technology has been selected as the basis for the baseline disruption mitigation system on ITER. The development of shattered pellet injection systems has progressed to be able to accelerate large pellets of pure argon and neon with or without including deuterium. Impact studies have been carried out at shallow angles to determine funnel performance in guiding pellets from multiple barrels into a common injection line and across pumping breaks. The characterization of the shattered spray has also progressed with fragment size measurements as a function of pellet speed showing a strong inverse relationship. Results of these studies are reported with implications for applications on existing and future tokamak devices.

Electrothermal mass accelerators, based on capillary discharges, that form a plasma propelling force from the ablation of a low-z liner material are candidates for fuelling magnetic fusion reactors. As lithium is considered a fusion fuel and not an impurity, lithium hydride and lithium deuteride can serve as good ablating liners for plasma formation in an electrothermal plasma source to propel fusion pellets. A comprehensive study of solid lithium hydride and deuteride as liner materials to generate a plasma to propel cryogenic fuel pellets is presented here. This study was conducted using the ETFLOW capillary discharge code. Relationships between propellants, source and barrel geometry, pellet volume and aspect ratio, and pellet velocity are determined for pellets ranging in volume from 5 to 100 mm3.

The realization of fusion energy requires materials that can withstand high heat and particle fluxes at the plasma material interface. The U.S. fusion energy program has placed development of such materials to be a top tier area of research interest. In this work, an electrothermal (ET) plasma source has been designed as a possible transient heat flux source for the Material Plasma Exposure eXperiment (MPEX) at Oak Ridge National Laboratory (ORNL). The ET source plasma can mimic transient events that impact the plasma facing components of a tokamak reactor, known as edge localized modes (ELMs). An ET plasma source operates in the ablative arc regime, which is driven by a DC capacitive discharge. The plasma arc current travels through the 4 mm bore of a liner and subsequently ablates and ionizes the liner material. This results in a high density plasma that carries a large bulk momentum out of the exit of the source.

Electrothermal plasma sources operating in the confined controlled arc discharge regime generate plasmas which can be used for a wide variety of applications, including materials deposition, mass acceleration devices, and high heat flux material exposure. Such plasmas are produced by capillary arc discharges-discharging high voltage across an insulated sleeve. Energy radiated to the walls of the insulated sleeve is absorbed by the material surface, ablating the material and forming a dense vapor. This vapor ionizes and forms a plasma which acts as a blackbody radiation source, which in turn can provide high heat flux to exposed surfaces. The generated plasma then exits the sleeve at high velocity. A substrate placed at the exit of the sleeve can act as a deposition surface for the ablated material.

In 1889, Friedrich Paschen published a law that describes the voltage at which a gas breaks down into plasma. It describes the dependence of the breakdown voltage on the gas pressure, distance between the electrodes, the secondary electron emission coefficient, and a set of proportionality constants [1]. A study by V. A. Lisovskiy et al [2] concluded that the original Paschen's law and the modification presented by Townsend in 1915 [3] were incorrect in assuming that breakdown voltage is dependent only on the product of the pressure times the inter-electrode distance (pd), and the coefficient of secondary electron emission (γs e ) from the surface of the cathode. The electron diffusion and mobility were considered in this modified breakdown condition [2]. Ledernez et al. has related the famous proportionality constants A and B of the voltage breakdown equation to the electron-ion collision cross sections for the working gas [4].

Pulsed electrothermal (ET) plasma sources were originally used as ignition sources for electrothermal chemical weapons and secondary rail gun armature. An ET system consists of a discharge capacitor and accompanying electronics. The capacitor discharges through an electrode, normally comprised of tungsten, which sends current to ground through an insulating liner. The liner can be any insulating material that is subject to dissociation and ablation due to joule heating.

The technology for producing, accelerating, and shattering large pellets (before injection into plasmas) for disruption mitigation has been under development at the Oak Ridge National Laboratory for several years, including a system on DIII-D that has been used to provide some significant experimental results. The original proof-of-principle testing was carried out using a pipe gun injector cooled by a cryogenic refrigerator (temperatures ∼8 to 20 K) and equipped with a stainless steel tube to produce 16.5 mm pellets composed of either pure D2, pure Ne, or a dual layer with a thin outer shell of D2 and core of Ne. Recently, significant progress has been made in the laboratory using that same pipe gun and a new injector that is an ITER test apparatus cooled with liquid helium. The new injector operates at ∼5 to 8 K, which is similar to temperatures expected with cooling provided by the flow of supercritical helium on ITER. An alternative technique for producing/solidifying large pellets directly from a pre-mixed gas has now been successfully tested in the laboratory. Also, two additional pellet sizes have been tested recently (nominal 24.4 and 34.0 mm diameters). With larger pellets, the number of injectors required for ITER disruption mitigation can be reduced, resulting in less cost and a smaller footprint for the hardware. An attractive option is longer pellets, and 24.4 mm pellets with a length/diameter ratio of ∼3 have been successfully tested. Since pellet speed is the key parameter in determining the response time of a shattered pellet system to a plasma disruption event, recent tests have concentrated on documenting the speeds with different hardware configurations and operating parameters; speeds of ∼100 to 800 m/s have been recorded. The data and results from laboratory testing are presented and discussed.

The realization of fusion energy requires materials that can withstand high heat and particle fluxes at the plasma material interface. In this work, an electrothermal (ET) plasma source has been designed as a transient heat flux source for a linear plasma material interaction device. An ET plasma source operates in the ablative arc regime driven by a DC capacitive discharge. The current channel width is defined by the 4 mm bore of a boron nitride liner. At large plasma currents, the arc impacts the liner wall, leading to high particle and heat fluxes to the liner material, which subsequently ablates and ionizes. This results in a high density plasma with a large unidirectional bulk flow out of the source exit. The pulse length for the ET source has been optimized using a pulse forming network to have durations of 1 and 2 ms. The peak currents and maximum source energies seen in this system are 1.9 kA and 1.2 kJ for the 2 ms pulse and 3.2 kA and 2.1 kJ for the 1 ms pulse, respectively. This work is a proof...

To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ∼1 GW/m2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured with two pulse lengths and tested under a solenoidal magnetic field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. The tungsten target plate is analyzed for surface damage using a scanning electron microscope.

Deep fueling for large-scale tokamak fusion reactors requires the use of high-velocity fuel pellet injectors. The fuel pellets consist of deuterium or deuterium-tritium ice. Electrothermal plasma guns can be used to launch fuel pellets into the fusion plasma at a range of velocities. The plasma guns can be used to launch pellets at the velocities required for fusion reactor deep fueling or for the control of edge localized modes.

Deep fueling for large-scale tokamak fusion reactors requires the use of high-velocity fuel pellet injectors. The fuel pellets consist of deuterium or deuterium-tritium ice. Electrothermal plasma guns can be used to launch fuel pellets into the fusion plasma at a range of velocities. The plasma guns can be used to launch pellets at the velocities required for fusion reactor deep fueling or for the control of edge localized modes. Electrothermal plasma guns use a capillary tube where plasma is sparked, called the source, and an acceleration barrel. The plasma is sparked inside the source using a capacitor bank that is charged to 10 kV. A liner material inside the source is ablated and forms plasma that draws up to 40 kA of current over 100 microseconds and can propel a pellet to velocities exceeding 3 km/s. Using a one dimensional computer code called ETFLOW, a variety of computational predictions can be made on the plasma as well as the pellet as it moves through the barrel. Using this code, different geometric configurations of the source have been simulated using an input current pulse with a current peak of approximately 20 kA. The geometries studied varied in source length and source radius. To compliment these studies, the same geometries have been simulated with different input current pulses. These input current pulses have maximum currents of 10 kA, 30 kA, and 40 kA. The results from all four current pulses are presented, and computed pellet velocities are correlated to source geometry and plasma parameters. Pellet exit velocities of up to 4.4 km/s were computed. This is an increase of 10% from previous computational studies.

In magnetic or inertial fusion reactors, hydrogen, its isotopes, and lithium are used as fusion fueling materials. Lithium is considered a fusion fuel and not an impurity in fusion reactors as it can be used to produce fusion energy and breed fusion products. Lithium hydride and lithium deuteride may serve as good ablating sleeves for plasma formation in an ablation-dominated electrothermal plasma source to propel fusion pellets. Previous studies have shown that pellet exit velocities greater than 3 km/s are possible using low-z propellant materials. In this work, a comprehensive study of solid lithium hydride and deuteride as a pellet propellant is conducted using the ETFLOW code. Relationships between propellants, source and barrel geometry, pellet volume and aspect ratio, and pellet velocity are determined for pellets ranging in volume from 1 to 100 mm3.

The electrothermal segmented plasma source `ETSPS' is a new concept in which the ablation of source capillary segments makes a mixed material plasma. In most ablation controlled capillary discharges, the inner ablating sleeve is usually made of an insulator with almost no use of metals, alloys, or semiconductors as most of the studies were devoted to the generation of high enthalpy plasma flows for launch technology using polyethylene or Lexan as the liner material. The ETFLOW code has been modified to include a materials library module for metals and alloys, structured amorphous metals, ceramics, and polymers. It has also been modified to allow for segmentation of the source with a different material in each segment to generate micro particles from the ablation of the segments, and the mixture of all of the ablated mass forms the final plasma at the source exit. The ETFLOW code is used to study the nature and composition of the plasma produced in the ETSPS to establish direction to experiment design. A set of computational experiments were conducted with a combination of silicon and graphite (carbon) to study the mixing ratios for SiC synthesis. Another set using molybdenum, sulfur and titanium segments was conducted to explore synthesis of MoSTi as a lubricant coating. The case study with SiC has shown that 7 carbon and 4 silicon segments can provide the desired ration for SiC synthesis (70% Si, 30%C). Another case study with 3 segments of molybdenum, 4 sulfur and 4 titanium segments has shown a plasma mixture of 57.00% sulfur, 27.62% molybdenum and 15.38% titanium, which provides the desired ratio of sulfur/molybdenum for MoSTi composition.