Hyper-Velocity Fullerene-Dusty Plasma Jets for Disruption Mitigation

High-velocity hydrogen plasma jets demonstrated deep penetration into magnetically confined tokamak plasma during fueling experiments. Disruption mitigation with plasma jets requires much larger mass (∼0.3 g for DIII-D and ∼30 g for ITER) to be delivered into the plasma core. We present a concept of using hyper-velocity C₆₀ fullerene-dusty plasma jets for disruption mitigation of magnetically confined plasmas. The C₆₀-fullerenes can form a compact plasma slug which can be accelerated to hyper-velocity (∼50 km/s) allowing for deep penetration into the tokamak target plasma. We evaluate the principle advantages, means to create it using a pulsed source, the atomic processes during acceleration to hyper-velocity and subsequent transport, and the interaction with the ambient target tokamak plasma during penetration.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

A Physics Exploratory Experiment on Plasma Liner Formation

Momentum flux for imploding a target plasma in magnetized target fusion (MTF) may be delivered by an array of plasma guns launching plasma jets that would merge to form an imploding plasma shell (liner). In this paper, we examine what would be a worthwhile experiment to explore the dynamics of merging plasma jets to form a plasma liner as a first step in establishing an experimental database for plasma-jets-driven magnetized target fusion (PJETS-MTF). Using past experience in fusion energy research as a model, we envisage a four-phase program to advance the art of PJETS-MTF to fusion breakeven (Q 1). The experiment (PLX) described in this paper serves as Phase 1 of this four-phase program. The logic underlying the selection of the experimental parameters is presented. The experiment consists of using 12 plasma guns arranged in a circle, launching plasma jets toward the center of a vacuum chamber. The velocity of the plasma jets chosen is 200 km/s, and each jet is to carry a mass of 0.2 mg to 0.4 mg. A candidate plasma accelerator for launching these jets consists of a coaxial plasma gun of the Marshall type.     

Effect of CH<Subscript>4</Subscript> Ion Implantation in Pure Aluminium

In this paper, aluminium samples with 99.96% purity were exposed to ion beam, extracted from CH₄ plasma. Implantation of ions were performed for 50 keV energy and various doses ranging from 1 × 10¹⁷ to 6 × 10¹⁷ ions/cm². Morphology of surfaces, roughness and its evolution during variation of ion dose has been studied by atomic force microscopy (AFM). Microstructure of the modified surfaces after ion implantation has been obtained by X-ray diffraction technique and Raman spectroscopy. Formation of aluminium carbide (Al₄C₃) was confirmed by XRD results at implantation doses of 3 × 10¹⁷ and 6 × 10¹⁷ ions/cm². In addition, it was observed that when the ion dose is increased, orientation of aluminium planes change from (2 2 0) to (2 0 0). Corrosion test was performed and compared for implanted and un-implanted samples. The results showed that corrosion resistivity increase by accumulation of ion dose.     

The Challenge of Wall–Plasma Interaction with Pulsed Megagauss Magnetic Fields

A method is described for choosing experimental parameters in studies of high-energy-density (HED) physics relevant to fusion energy, as well as other applications. An important HED issue for magneto-inertial fusion (MIF) is the interaction of metal pusher materials with megagauss (MG) magnetic fields during liner compression of magnetic flux and fusion fuel. The experimental approach described here is to study a stationary conductor when a pulsed current generates MG fields at the surface, instead of studying the inner surface of a moving liner. This places less demand upon the pulsed power system, and significantly improves diagnostic access. Thus the deceptively simple geometry chosen for this work is that of a z pinch composed of a metal cylinder carrying large current. Consideration of well known stability issues for the z pinch shows that for given peak current and rise time from a particular power supply, there is a minimum radius and thus maximum B field that can be created without disruption of the conductor before peak current. The reasons are reviewed why MG levels of magnetic field, as required for MIF, result in high temperatures and plasma formation at the surface of the metal in response to Ohmic heating. The distinction is noted between the liner regime obtained with cylindrical rods, which have a skin depth small compared to the conductor radius, and the exploding thin-wire regime, which has skin depth larger than the wire radius. A means of diagnostic development is described using a small facility (DPM15) built at the University of Nevada, Reno. It is argued that surface plasma temperature measurements in the 10-eV range are feasible based on the intensity of visible light emission.     

Results of U<Subscript>0.55</Subscript>Pu<Subscript>0.45</Subscript>N and U<Subscript>0.4</Subscript>Pu<Subscript>0.6</Subscript>N mixed mononitride fuel tests in a bor-60 reactor to burnup 12% h.a.

The results of post-reactor studies of U_0.55Pu_0.45N and U_0.4Pu_0.6N mixed mononitride fuel elements (density 85% of the theoretical value) and a helium sublayer are presented. The fuel elements are irradiated in a BOR-60 reactor to burnup 9.4 and 12.1% h.a., respectively, with power density 430 and 540 W/cm. All fuel elements remained hermetic; the ChS-68 steel cladding (20% cold deformation) retained excess plasticity. The maximum zone of interaction between the cladding and the fuel and fission products did not exceed 15 μm. The swelling rate of U_0.4Pu_0.6N and U_0.55Pu_0.45N fuel was 1.1 and 0.68%/% burnup, respectively. The gas release did not exceed 19.3 and 19%. The steel damage dose was 43 dpa. The character of the porosity distribution in the fuel affects the swelling and gas release.     

Physics Criteria for a Subscale Plasma Liner Experiment

Spherically imploding plasma liners, formed by merging hypersonic plasma jets, are a proposed standoff driver to compress magnetized target plasmas to fusion conditions (Hsu et al. in IEEE Trans Plasma Sci 40:1287, 2012). In this paper, the parameter space and physics criteria are identified for a subscale, plasma-liner-formation experiment to provide data, e.g., on liner ram-pressure scaling and uniformity, that are relevant for addressing scientific issues of full-scale plasma liners required to achieve fusion conditions. Based on these criteria, we quantitatively estimate the minimum liner kinetic energy and mass needed, which informed the design of a subscale plasma liner experiment now under development.     

Technical Summary of the First U.S. Plasma Jet Workshop

This paper provides a technical summary of the first U.S. Plasma Jet Workshop, which was sponsored by the DOE Office of Fusion Energy Sciences and held at Los Alamos National Laboratory on January 24–25, 2008. The purpose of the workshop was to bring together members of the national plasma jet research community in order to discuss ongoing research and identify research needs and opportunities in plasma jets and their applications, which include fundamental studies of high energy density (HED) plasmas, magneto-inertial fusion (MIF), laboratory astrophysics, and disruption mitigation and fueling for magnetic confinement devices. Over the course of the workshop, about equal time was devoted to short technical talks and group discussions.     

Compression, Heating and Fusion of Colliding Plasmoids by a Z-theta Driven Plasma Liner

A potentially promising approach to fusion employs a plasma shell to radially compress two colliding plasmoids. The presence of the magnetic field in the target plasma suppresses the thermal transport to the confining shell, thus lowering the imploding power needed to compress the target to fusion conditions. With the momentum flux being delivered by an imploding plasma shell, many of the difficulties encountered in imploding a solid metal liner are eliminated or minimized. The best plasma for the target in this approach is the FRC. It has demonstrated both high β, and robustness in translation and compression that is demanded for the target plasma. A high density compressed plasmoid is formed by a staged axial and radial compression of two colliding/merging FRCs where the energy that is required for the implosion compression and heating of the magnetized target plasmoid is stored in the kinetic energy of the plasmas used to compress it. An experimental apparatus is being constructed for the demonstration of both the target plasmoid formation as well as the compression of the plasmoid by a plasma liner. It is believed that with the confinement properties and the high β nature of the FRC, combined with the unique approach to be taken, that an nτ_E T _i triple product ∼5 × 10¹⁷ m⁻³ s keV can be achieved.     

Recent progress on the Tokamak Fusion Test Reactor

The deuterium-tritium (D-T) experiments on the Tokamak Fusion Test Reactor (TFTR) have yielded unique information on the confinement, heating and alpha particle physics of reactor scale D-T plasmas as well as the first experience with tritium handling and D-T neutron activation in an experimental environment. The D-T plasmas produced and studied in TFTR have peak fusion power of 10.7 MW with central fusion power densities of 2.8 MWm^–3 which is similar to the 1.7 MWm^–3 fusion power densities projected for 1,500 MW operation of the International Thermonuclear Experimental Reactor (ITER). Detailed alpha particle measurements have confirmed alpha confinement and heating of the D-T plasma by alpha particles as expected. Reversed shear, highl _i and internal barrier advanced tokamak operating modes have been produced in TFTR which have the potential to double the fusion power to 20 MW which would also allow the study of alpha particle effects under conditions very similar to those projected for ITER. TFTR is also investigating two new innovations, alpha channeling and controlled transport barriers, which have the potential to significantly improve the standard advanced tokamak.     

The Role of Magnetized Liner Inertial Fusion as a Pathway to Fusion Energy

We discuss the possible impacts of a new magnetized liner inertial fusion concept on magneto-inertial fusion approaches to fusion energy. Experiments in the last 1.5 years have already shown direct evidence of magnetic flux compression, a highly magnetized fusing fuel, significant compressional heating, a compressed cylindrical fusing plasma, and significant fusion yield. While these exciting results demonstrate several key principles behind magneto-inertial fusion, more work in the coming years will be needed to demonstrate that such targets can scale to ignition and high yield. We argue that justifying significant investment in pulsed inertial fusion energy beyond target development should require well-understood, significant fusion yields to be demonstrated in single-shot experiments. We also caution that even once target ideas and fusion power plants have been demonstrated, historical trends suggest it would still be decades before fusion could materially impact worldwide energy production.     

A study of fullerite ablation with energetic pulsed electrons

We report the ablation of fullerite films deposited on metallic substrates with 3 keV electron pulses generated in a specially designed pulsed discharge tube. During ablation the fragmented species were detected by emission spectroscopy. The emission spectra of C₂ and C₁^∗ (CII) provide the signatures of C₆₀ fragmentation. The vibrational temperature of the C₂ emitted from the ablated fullerite is ∼12,700 ± 1160 K compared with ∼18,230 ± 1150 K for the graphite sample under similar conditions. The fullerite films were produced by vacuum sublimation on Aluminum, Iron and Copper substrates and characterized by Atomic force microscope, X-ray diffraction; Raman and Fourier transform infrared spectroscopy. The comparisons of electron ablation of fullerite films with that of graphite show the similarities and differences of carbon bonding in the caged structure of C₆₀ with that of the planar graphene sheets of graphite.     

The Plasma Liner Compression Experiment

Magnetic fusion at high energy density requires compression of a magnetized target in a manner that can be repetitively performed. One approach being investigated in these experiments is the radial compression of a FRC target by a cylindrical plasma liner. A circular array consisting of 16 small coaxial plasma sources were placed at each end of a quartz chamber radially near the wall to form the plasma liner. Liner masses of 3 mg have been formed in Xenon and have been accelerated to a velocity of 50 km/s. An FRC was produced in an adjacent chamber and translated inside the plasma liner where theta-pinch coils were employed to drive the plasma liner radially inward compressing the FRC. Close agreement is found between experimental results and 2D MHD numerical calculations. Efficient coupling (~50%) between the compression bank and the plasma liner has been observed.     

Numerical and Statistical Analysis of FR Spent Fuel Transmutation in a Thorium Fusion Breeder

In this study, a numerical analysis and an analysis of variance (ANOVA) are applied to find the best suitable neutronic parameters for the performance analysis in a thorium fusion rector. The numerical and ANOVA approach are employed to investigate the neutronic characteristics of a fusion reactor using ThO₂ 90% + FR spent fuel 10% fuel mixtures. Three different neutronic parameters for the ANOVA and numerical approach, namely, moderator/fuel volume fractions (V_m/V_f), plasma chamber dimensions (PCD) and neutron wall loading (NWLs) as time dependent are selected for neutronic performance characteristics including tritium breeding ratio (TBR), multiplication factor (M), total fission rate (Σ_f), ²³²Th(n,γ) reaction, burn up and/or transmutation (B/T) and fissile fuel breeding (FFBR). Moreover, effects of the NWLs, V_m/V_f fractions and PCD in the B/T of FR spent fuel mixed thorium are investigated. Numerical and statistics approach results are evaluated for TBR, M, Σ_f fission rate, ²³²Th(n,γ) reaction, B/T and FFBR.     

On the proliferation issues of a fusion fission fuel factory using a molten salt

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF₄ and UF₄ solute salts are transported by a LiF-BeF₂ solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle.The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S.For the Th-U cycle the bred material output remains below 10 g/h with a ²³²U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.     

Could Advanced Fusion Fuels Be Used with Today's Technology?

Could today's technology suffice for engineering advanced-fuel, magnetic-fusion power plants, thus making fusion development primarily a physics problem? Such a path would almost certainly cost far less than the present D-T development program, which is driven by daunting engineering challenges as well as physics questions. Advanced fusion fuels, in contrast to D-T fuel, produce a smaller fraction of the fusion power as neutrons but have lower fusion reactivity, leading to a trade-off between engineering and physics. This paper examines the critical fusion engineering issues and related technologies with an eye to their application in tokamak and alternate-concept D-³He power plants. These issues include plasma power balance, magnets, surface heat flux, input power, fuel source, radiation damage, radioactive waste disposal, and nuclear proliferation.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

The Plasma Jet/Laser Driven Compression of Compact Plasmoids to Fusion Conditions

Magneto-inertial fusion (MIF) is based on both magnetic and inertial confinement. An embedded magnetic field is compressed along with the target plasma to achieve magnetic insulation and fusion condition. Several magnetic systems for plasma confinement may be used for laser-driven (LD) and plasma jet driven (PJ) magnetic flux compression. Estimations show the possibility in principle to realize regimes of PJMIF system with a plasma gain factor Q > 10.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

Beam plasma neutron sources based on beam-driven mirror

The design and performance of a relatively low-cost, plasma-based, 14-MeV D-T neutron source for accelerated end-of-life testing of fusion reactor materials are described in this article. An intense flux (up to 5×10¹⁸ n/m²·s) of 14-MeV neutrons is produced in a fully-ionized high-density tritium target (n _e 3×10²¹ m^–3) by injecting a current of 150-keV deuterium atoms. The tritium plasma target and the energetic D⁺ density produced by D⁰ injection are confined in a column of diameter 0.16 m by a linear magnet set, which provides magnetic fields up to 12 T. Energy deposited by transverse injection of neutral beams at the midpoint of the column is conducted along the plasma column to the end regions. Longitudinal plasma pressure in the column is balanced by neutral gas pressure in the end tanks. The target plasma temperature is about 200 eV at the beam-injection position and falls to 5 eV or less in the end region. Ions reach the walls with energies below the sputtering threshold, and the wall temperature is maintained below 740 K by conventional cooling technology.