Development of Merged Compact Toroids for Use as a Magnetized Target Fusion Plasma

We report on the development of compact toroid (CT) accelerators to create the target plasma for magnetized target fusion (MTF) devices. Due to the requirements of high initial density of ~10¹⁷ cm⁻³, strong internal fields of 5–10 T, and base temperatures of >100 eV, a design based on conical compression electrodes is an effective avenue to pursue. Progress is being made at General Fusion Inc, (Vancouver, Canada) to develop a pair of large CT accelerators for generating an MTF target plasma. In this design, tungsten coated conical electrodes (with a formation diameter of 1.9 m, a radial compression factor of 4, and overall accelerator length of 5 m) will be used to achieve ohmic heating and acceleration of the CT, yet with low wall sputtering rates. A pair of these accelerators can be synchronized and shot at one another, producing a collision and reconnection of the two CTs within the center of an MTF chamber. Depending on the choice of relative helicities, the two CTs will merge to form either a spheromak-like or an FRC-like plasma. ICC 2008 Reno NV, June 25th, IP: 021.     

C<Subscript>60</Subscript>-Fullerene Hyper-Velocity High-Density Plasma Jets for MIF and Disruption Mitigation

We present an innovative idea to use hyper-velocity (>30 km/s) high-density (>10¹⁷ cm⁻³) plasma jets of D-T/H and C₆₀-fullerene for magneto-inertial fusion (MIF), high energy density laboratory plasma (HEDLP), and disruption mitigation in magnetic fusion plasma devices. The mass (~1–2 g) of sublimated C₆₀ and hydrogen (or D-T fuel) produced in a pulsed power source is ionized and accelerated as a plasma slug in a coaxial plasma accelerator. For MIF/HEDLP we propose to create a magnetized plasma target by injecting two high-Mach number high-density jets with fuel (D-T) and liner (C₆₀/C) structure along the axis of a pulsed magnetic mirror. The magnetized target fusion (MTF) plasma created by head-on collision and stagnation of jets is compressed radially by a metallic liner (Z-pinch) and axially by the C₆₀/C liner. For disruption mitigation, the C₆₀ plasma jets were shown to be able to provide the required impurity mass (J Fusion Energy 27:6, 2008).     

Heliotron E Results and Large Helical Project

Heliotron E(H-E) experiment was started in 1980. Until 1987 high power heating experiments for improving plasma parameters have almost finished. H-E firstly demonstrated that ECR heated plasmas are usable for target plasmas of NBI or ICRF heating to obtain high density and high temperature currentless plasmas. The highest electron temperature is 1.5keV and ion temperature is 1.6keV and both are realized in the low density regime of <n> (average density) ≤10¹³cm^?3.

H-E also showed that the currentless plasmas have no major disruption and quasi-steady plasmas are confined with controlling impurity ions by titanium gettering and carbon coating.

H-E also obtained <β> (average β) –2%, which is the highest value realized in helical systems, with <n–8×l0¹³cm^?3 and Te(0)–T_i(0)–350 eV at B₀ (magnetic field at the magnetic axis) =0.94 T. In the high β experiments pressure-driven instabilities were observed for peaked pressure profiles and sometimes relaxation oscillations similar to the tokamak internal disruptions were observed.

In the ECRH plasmas neoclassical transport is dominant in the region inside the half radius. However, global confinement time τ_E follows the scaling law τ_E ∝<n>^0.66P_heat ^?0.53 which is different from the neoclassical scaling law. Here P_heat denotes the net heating power.

Based on the H-E results, a new large helical system design study has started in 1986. The plasma parameters entering the regime of <n>τ_E<T> (2–3)× 10¹⁹m^?3?S?keV is investigated, which is about one tenth of fusion plasma condition. From the transport code studies and empirical scaling law based on the H-E results, R=(4×5)m, ā=(50–60)cm and B_o=4T are required to satisfy the above condition with P_heat=20MW. The design study to fix the magnetic field configuration is progressing. Expected one is l=2 and m=10 with additional poloidal coils, where m is a toroidal period number. The magnetic field is produced by superconducting coil and long pulse operation will be tested, if continuous heating is available.     

Monte Carlo studies of charge-exchange first wall heat fluxes from neutral beam injection in tandem mirror reactors

Monte Carlo methods are employed to compute surface heat fluxes at the first walls of three representative tandem mirror reactors: TDF, TASKA-M, and MARS, resulting from charge-exchange (CX) of high power neutral beams with the fusion plasmas. The full three-dimensional nature of the anisotropic interaction processes of neutral beams with mirror-confined plasmas is retained in the calculations, and resulting CX heat fluxes are mapped both azimuthally and axially on the first wall surfaces. The angular distribution of heat fluxes at the first wall shows strong backward-peaking for TDF with maximum power densities of ～ 2.4 kW cm^?2 occurring at 180° relative to the incident beam direction. By contrast, the angular distributions for TASKA-M and MARS exhibit strong forward-peaking, resulting in hot spots of 1140 and 700 W cm^?2, respectively, at the first wall. Physical arguments are presented for the behavior of these CX heat flux distributions in terms of the plasma and beam parameters of each system. Particular emphasis is given to the engineering implications of the results and methods for ameliorating these high first wall heat fluxes are discussed.     

Material testing in a linear theta pinch

The interaction of stainless steel 316 and Inconel 625 alloys has been investigated with a thermonuclear-like plasma, n = 10¹⁶cm^?3 and T_i = 1 keV, generated in the Alvand I linear theta pinch. The average power flux is 10⁷ W/cm² and the interaction time nearly one μs. A theoretical analysis based on the formation of an observed impurity layer near the material, has been used to determine the properties of the impurity layer and the extent of the damage on the material. Although arcing has been observed, the dominant damage mechanism has been assessed to be due to evaporation. Exposure to single shots has produced very heavily defective areas and even surface cracks on the SS 316 sample, but no cracks were observed on Inconel 625 after exposure to even 18 shots. On the basis of temperature rise and evaporation a comparison is made among materials exposed to plasmas of a theta pinch, shock tube, present generation tokamak and an anticipated tokamak reactor.     

Fluxes of charged and neutral particles from tokamaks

The performance of present tokamak devices is significantly influenced by neutral atom and impurity ion populations originating at the wall or aperture limiter. Plasma instability and diffusion mechanisms cannot be identified until these surface-related processes are accurately described. Central neutral densities in the Oak Ridge tokamak vary from 2 × 10⁸ cm^?3 to 2 × 10⁹ cm^?3 and agree well with values predicted by our theoretical model. We extrapolate this model to the larger, denser and hotter plasmas foreseen in the next generation of experiments and beyond. We calculate the rate of impurity generation by charge exchange bombardment. Impurities will influence scientific feasibility experiments which burn D-T by raising ignition temperatures. In a version of our code which incorporates alpha heating, bremsstrahlung, synchrotron and line radiation we calculate the effects of small admixtures of high $\text{Z}$ ions on the injection power required to reach ignition and on the required pulse duration. We give the wall energy fluxes to the expected and the relative fractions of energy conduction and radiation.     

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.     

Simulation of Electron Trajectories in the Multicusp Ion Source Using Geantn4 Monte Carlo Code

To optimize the negative ion source and generate intense beams of negative ions, understanding of transport properties of both electrons and negative ions is indispensable. Transport process of negative hydrogen ions (H⁻) in a multicusp H⁻ source, has been simulated by three-dimensional Femlab simulation software. Multipolar plasma confinement is known to result in enhanced plasma density, homogeneous plasma of a large volume, and quiescent plasmas. The effect of plasma confinement by applying multi-polar magnetic field was investigated. Results are obtained for ten different configurations of permanent magnet and discussed. Full line cusps are found to give optimum plasma density. Negative ions created on the sidewall hardly can reach the center of the source due to trapping by the multicusp magnetic field. As a result, H⁻ ions created on the sidewall do not have a significant effect on the H⁻ current.     

Perspectives on Magnetized Target Fusion Power Plants

One approach to Magnetized Target Fusion (MTF) builds upon the ongoing experimental effort (FRX-L) to generate a Field Reversed Configuration (FRC) target plasma suitable for translation and cylindrical-liner (i.e., converging flux conserver) implosion. Numerical modeling is underway to elucidate key performance drivers for possible future power-plant extrapolations. The fusion gain, Q (ratio of DT fusion yield to the sum of initial liner kinetic energy plus plasma formation energy), sets the power-plant duty cycle for a nominal design electric power [e.g. 1,000 MWe(net)]. A pulsed MTF power plant of this type derives from the historic Fast Liner Reactor (FLR) concept and shares attributes with the recent Inertial Fusion Energy (IFE) Z-pinch and laser-driven pellet HYLIFE-II conceptual designs. Work supported by the Office of Science, OFES, through Los Alamos National Laboratory, under DOE contract W-7405-ENG-36.     

Light ion bombardment sputtering,stress buildup,and enhanced surface contamination

Several effects of high fluence light ion bombardment which are relevant to fusion reactor inner wall problems are under investigation. The impurity loading and lateral stress from high fluence ion bombardments can alter sputtering yields markedly. This is demonstrated with sputtering measurements of 45 keV Kr incident upon Au. Sputtering yields for 150 keV⁴ He onto Au are presented for two different background hydrocarbon partial pressures during bombardment. It is shown that there is a polymerized hydrocarbon buildup on the surface for hydrocarbon partial pressures greater than $\text{1 × 10}{}^{\mathrm{?10}}\text{mm Hg}$ even for ion current densities ranging as high as 4 μA/cm²; true sputtering of the Au has been observed only for lower hydrocarbon partial pressures. Additional effects of oxidizing background gases on sputtering measurements of reactive metals are discussed.It is concluded that the background gas partial pressures, temperatures, and particle fluxes in a fusion device and in simulation experiments must be well defined before sputtering effects can be understood.     

Radiative Cooling Rates for Low-Z Impurities in Non-coronal Equilibrium State

The local or transient radiation losses in tokamak plasmas can greatly exceed those in the coronal equilibrium. This excess is especially pronounced at the plasma edge. The reason for the increase of radiation in a peripheral plasma is as follows. The impurities are lost fast from the plasma edge and the new impurity source is supplied to this region. The charged states of impurities, therefore, do not reach their coronal equilibrium ones. These impurity ions have more electrons than those in the coronal equilibrium, and as a result emit the higher radiation power. In the simplest case, the non-coronal radiative rate can be determined only by two parameters: the electron temperature \(T_{\text {e}}\) and the so-called “residence parameter” \(n_{\text {e}}\tau _{\text {i}}\), where \(\tau _{\text {i}}\) is the impurity residence time in the plasma. Despite the strong simplification, such an approach allows to do simple estimates of non-coronal radiation. In this paper, two dimensional polynomial fits describing radiative cooling rates and mean charge are obtained for eight impurity species: helium, lithium, beryllium, carbon, nitrogen, oxygen, neon, and argon. The results are presented in figures and tables. The figures show curves calculated from the original atomic database and least-squares polynomial fits to these curves. The tables contains coefficients for this fits. The obtained fits can be useful for qualitative estimates and simple numerical calculations.     

In search of nuclear fusion in electrolytic cells and in metal/gas systems

It has been reported recently in the literature that unexpected thermal and nuclear effects (production of excess heat, neutrons, γ-rays, and tritium) can occur during the electrolysis of heavy water at palladium or titanium electrodes, or during temperature and pressure cycling of the titanium/deuterium gas system. We have attempted to reproduce some of these experiments. A variety of electrochemical cells having palladium cathodes in the form of wires, tubes, sheets, and rods have been used to electrolyze heavy water containing 0.1 mol.dm⁻³ LiOH, 0.1 mol.dn⁻³ LiOD or 0.5 mol.dm⁻³ D₃PO₄. Current densities of up to 200 mA.cm⁻² were applied. The mass of the palladium cathodes covered the range from 1–40 grams and the surface area varied from 8–140 cm². Neutron detection systems with low constant backgrounds were used to search for neutron emission during electrolysis. These included³He- and¹⁰BF₃-based detectors. After running some of the cells for more than 30 days, no neutron emission above background could be detected. This puts upper limits of 0.5 s⁻¹ and 2×10⁻²³ fus. D-D.s⁻¹ on the neutron emission and the fusion rate, respectively. A sensitive and accurate heat-flow calorimeter was built and used to monitor the energy balance of some of the cells during electrolysis. No unexpected heat effects were observed. This puts an upper limit of 0.13 W.cm⁻³ on the specific excess power. No enrichment of the electrolyte in tritium was evident after electrolysis. Experiments were also performed with the titanium/ deuterium gas system. These consisted of exposing titanium metal to a deuterium gas pressure of 40 atmospheres, lowering the temperature to −196°C, releasing the pressure and gradually warming the titanium to room temperature. No neutron emission above background was observed during these experiments, which puts upper limits of 0.5 s⁻¹ and 4×10⁻²⁵ fus.D-D.s⁻¹ on the neutron emission and fusion rate, respectively. Submitted toJournal of Fusion Energy as part of the Proceedings of the Workshop on Cold Fusion Phenomena held in Santa Fe in May 1989.     

Observation of Central Toroidal Rotation Induced by ICRF on EAST

Core plasma rotation of both L-mode and H-mode discharges with ion cyclotron range of frequency(ICRF) minority heating(MH) scheme was measured with a tangential X-ray imaging crystal spectrometer on EAST(Experimental Advanced Superconducting Tokamak).Cocurrent central impurity toroidal rotation change was observed in ICRF-heated L-and H-mode plasmas.Rotation increment as high as 30 km/s was generated at ~1.7 MW ICRF power.Scaling results showed similar trend as the Rice scaling but with significant scattering,especially in L-mode plasmas.We varied the plasma current,toroidal field and magnetic configuration individually to study their effect on L-mode plasma rotation,while keeping the other major plasma parameters and heating unchanged during the scanning.It was found that larger plasma current could induce plasma rotation more efficiently.A scan of the toroidal magnetic field indicated that the largest rotation was obtained for on-axis ICRF heating.A comparison between lower-single-null(LSN)and double-null(DN) configurations showed that LSN discharges rendered a larger rotation change for the same power input and plasma parameters.     

Studies of the Hard X-ray Emission from the Filippov Type Plasma Focus Device, Dena

This article is about the characteristics of the hard X-ray (HXR) emission from the Filippov type plasma focus (PF) device, Dena. The article begins with a brief presentation of Dena, and the mechanism of the HXR production in PF devices. Then using the differential absorption spectrometry, the energy resolved spectrum of the HXR emission from a 37 kJ discharge in Dena, is estimated. The energy flux density and the energy fluence of this emission have also been calculated to be 1.9 kJ cm⁻² s⁻¹ and 9.4 × 10⁻⁵ J cm⁻². In the end, after presentation of radiography of sheep bones and calf ribs, the medical application of the PF devices has been discussed.     

Confinement of Strongly Anisotropic Hot-ion Plasma in a Compact Mirror

The paper presents the results of recent study of anisotropic plasma with thermonuclear ions confined in the axially symmetric Gas Dynamic Trap (GDT) mirror. Anisotropic ions are produced by the perpendicular injection of two focused 18 keV neutral beams in the small mirror section attached to the GDT central cell. We observed build-up of density of anisotropic ions up to approximately 10¹³cm⁻³ with the localized spatial profile and the mean energy of 7 keV. The average fast ion density three times exceeded the background plasma density. Fast ion accumulation is accompanied by the decrease of the plasma flux from the central cell recorded outside the mirror, that was qualified as a potential barrier development causing confinement improvement. Analysis of measurement results compared with estimates of plasma parameters in the compact mirror allows to scale to the experiments with next generation neutral beams with increased power and pulse duration.     

Operation of VVR-SM with 36% Enrichment Uranium-Dioxide Fuel

VVR-SM after conversion to IRT-3M fuel assemblies with 36% fuel enrichment is operating at 10 MW with a core load of 18 fuel assemblies. The safety of the operation is ensured by neutrons and thermohydraulic calculations of the operating regimes of the core after each reloading of fuel assemblies. The experimental channels of nine fuel assemblies with flux density of neutrons with energy >0.821 MeV up to 1.1⋅10¹⁴ sec⁻¹⋅cm⁻² are used for producing ³²P and ³³P. The flux density of thermal neutrons in the channels of the beryllium reflector reaches 1.6⋅10¹⁴ sec⁻¹⋅cm⁻². Nine of the horizontal channels and the channel of the thermal column are used for fundamental and applied studies and for neutron-activation analysis. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 147–152, August 2005.     

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.     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

Dense Plasma Injection Experiment at MCX

We present preliminary results of the High Density Plasma Injection Experiment at the Maryland Centrifugal Experiment (MCX). HyperV Technologies Corp. has designed, built, and installed a prototype coaxial gun to drive rotation in MCX. This gun has been designed to avoid the blow-by instability via a combination of electrode shaping and a tailored plasma armature. An array of diagnostics indicates the gun is capable of plasma jets with a mass of 160 μg at 70 km/s with an average plasma density above 10¹⁵ cm⁻³. Preliminary measurements are underway at MCX to understand the penetration of the plasma jet through the MCX magnetic field and the momentum transfer from the jet to the MCX plasma. Data will be presented for a wide range of MCX field parameters, and the prospects for future injection experiments will be evaluated.     

Experimental Results for an Acoustic Driver for MTF

General Fusion is planning to form an FRC or spheromak of 10¹⁷ cm⁻³, 100 eV, 40 cm diameter by merging two spheromaks with reverse or co-helicity. This target will be further compressed in a 3 m diameter tank filled with liquid PbLi with the plasma in the center. The tank is surrounded with pneumatically powered impact pistons that will send a convergent shock wave in the liquid to compress the plasma to 10²⁰ cm⁻³, 10 keV, 4 cm diameter for 7 μs. General Fusion has built a 500 kJ, 80 μs, 6 GW pneumatic impact piston capable of developing 2 GPa (300 kpsi). In this paper we will present the performances achieved to date.