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.     

Optimization of Fusion Pellet Launch Velocity in an Electrothermal Mass Accelerator

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.     

Flow Characteristics and Charge Exchange Effects in a Two-Dimensional Model of Electrothermal Plasma Discharges

Electrothermal (ET) plasma discharges are capillary discharges that ablate liner materials and form partially ionized plasma. ET plasma discharges are generated by driving current pulses through a capillary source with peak currents on the order of tens of kA and pulse lengths on the order of \(100\,\upmu \hbox {s}\). These plasma discharges can be used to propel pellets into magnetic confinement fusion devices for deep fueling of the fusion reaction, ELM mitigation, and thermal quench of the fusion plasma. ET plasma discharges have been studied using 0D, 1D, and semi-2D fluid models. In this work, a fully 2D model of ET plasma discharges is presented. The newly developed model and code resolve inter-species interaction forces due to elastic collisions. These forces affect the plasma flow field in the source and impede the development of plasma pressure at the exit of the source. In this work, these affects are observed for discharge current pulses peaking at 10 and 20 kA. The sensitivity of the model to the inclusion of charge exchange effects is observed. The inclusion of charge exchange has little effect on the integrated, global results of the simulation. The difference in total ablated mass for the simulations caused by the inclusion of charge exchange reactions is <1 %. Differences in local plasma parameters are observed during discharge initialization, but after initialization, these differences diminish. The physical reasoning for this is discussed and recommendations are made for future modeling efforts.     

Innovative Fusion Tokamak Powerplant and the Basic Engineering for Mhd Theory

All around the world an endeavour to develop the fusion process as a major alternative energy has been going on for about a half century. Aries-St is the spherical tokamak (St) a innovative fusion reactor engineering. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together (Tillack et al. in Fusion Energ Des 65:215–261, 2003; El-Guebaly in Fusion Energ Des 65:263–284, 2003). The Aries-St power core consist of the components directly surrounding the burning plasma and serves important functions. In fusion applications, liquid metals are traditionally considered to be the best working fluids. Sufficient tritium breed amount must be TBR >1.1 for Aries-St fusion tokamak power plant (Tillack et al. in Fusion Energ Des 65:215–261, 2003; El-Guebaly in Fusion Energ Des 65:263–284, 2003). The Aries-St power core has designed for correlation with an optimized St plasma that develop through the investigation of extensive range of plasma magnetohydrodynamic (Mhd) equations. In this study, the engineering design plasma parameters are described with respect to Mhd equilibrium and nuclear analysis, stability, radiation heat transfer conditions, current drive, and safety. In addition, turbulence model extended to an incompressible Mhd flows and monte carlo simulation are used for modeling of low-conductivity fluid. In this study the modeling of aries-st tokamak reactor produced by using aries design technology, has performed by using the monte carlo code and Endf/b-V-VI nuclear data. Monte carlo method is the general name for the solution of experimental and statistical problems with a random approach.     

The Effect of Plasma Profiles on the Critical Value of $$n\uptau_{E}$$ for Ignition

The effect of plasma profiles for ignition condition in a stationary D–T plasma is investigated using the energy conservation equations for ions and electrons, assuming that steady state fusion power is produced with no external power. The alpha power heating is sufficiently large to sustain the plasma and to balance the combined Bremsstarhlung and thermal conduction losses. The space dependent Lawson criteria is derived and critical condition is identified. As a result of this analysis we have shown that the optimum temperature might be \(\bar{T} \approx 26\,{\text{keV}}\) and that the peaked profiles with \(n\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{n} }}\), ν _n  = 1, and \(T\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{T} }} ,\,v_{T} = 2\) are good to minimizing \(\bar{n}\uptau_{E}\) for ignition. The results for these profiles show the critical value of \((\bar{n}\uptau_{E} )_{min} = 0.08 \times 10^{20 } \,{\text{m}}^{ - 3} \,{\text{s}}\) showing the reduction by 1/3 from the reference value limit ν _n  = ν _T  = 0. For a 26 keV plasma with an energy confinement time of 1 s, a pressure of about 6.24 atm is required for the plasma to be ignited; that is, it is sustained purely by the self-heating of the fusion alpha particles.     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).     

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.     

Interaction of imploding shock waves with expanding central plasma in spherical pinch experiments: Simulation analysis

In the spherical pinch scheme, the hot D-T plasma produced in the center of the high pressure spherical vessel is confined by means of imploding shock waves launched from the periphery of the vessel for a time sufficiently long to achieve break-even conditions for plasma fusion. Theoretical studies on spherical pinch made so far have been limited up to the conditions of substantial expansion of the central plasma and the well-defined time delay between the creation of central plasma and the launching of the peripheral shock which led to the conclusion that, in realistic situations of SP experiments, negative time delays should be adopted, i.e., the launching of the imploding shock wave should precede the formation of the central plasma. However, the interaction of converging shock wave with the central plasma causing an additional heating and compression of the central plasma favoring plasma fusion conditions was not taken into account. Starting from the hydrodynamic equations of the system, the proposed simulation code deals with the propagation of converging shock waves and its interaction with the expanding central plasma. Considering the above-mentioned interaction in a self-consistent manner, the temporal evolution of temperature of central plasma is studied. Some results of the numerical simulation on the dynamics of shock wave propagation are also compared with the predictions of point strong explosing theory.     

Power and Sizes of Tokamak Fusion Neutron Sources with NBI-Enhanced Reaction Rate

Steady state tokamak with deuterium–tritium plasma is considered as a basis for fusion neutron source for a hybrid fusion–fission reactor. Prototypes of such a system can be developed on the basis of the present day tokamaks as the plasma power gain factor Q ~ 1 is required for hybrid applications. Significant population of fast ions can be supported by a powerful neutral beam injection heating in regimes with Q ~ 1. The reaction rate for fast ions greatly exceeds the rate for thermal Maxwellian ions. The possible ranges of parameters are discussed for medium size tokamaks with minor plasma radius a = 0.5–1 m. Power and sizes of the neutron source are determined by the value of the injection energy. Power gain Q ≈ 1 can be achieved with injection energy of deuterium about 130 keV and tritium energy about 200 keV. Neutron power of 30–40 MW can be realized with a ≈ 1 m, and about of few megawatts with a ≈ 0.5 m.     

Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.     

Exploratory studies of tokamaks as fusion test reactors

Studies have been performed to explore various plasma burn scenarios for a tokamak test reactor which could follow the next generation of large tokamak experiments. Tradeoffs between an ignited burning plasma and a sub-ignited driven plasma are examined in terms of device size and performance as a fusion engineering test facility. It is found that plasma performance levels, measured by ignition margin, amplification factorQ, and fusion power output, increase with device size, more optimistic transport scaling laws, lower magnetic field ripple, and higher. The performance of a generally low stress (B ₀=4 T) reference device, with major radiusR=4.5 m and minor radiusa=1.3 m in a D-shaped (=1.6) plasma has been evaluated over a wide range of operating parameters. In particular, a moderate fusion power output of 300 MW is obtained, the driven plasma havingQ 10, an edge ripple of 1%, and a density ranging between 1.0 and 1.5×10¹⁴ cm^–3. The same device operated at a higher general level of stress (B ₀=5.3 T) is predicted to achieve ignition, but is not required for the mission of an engineering test facility and would entail greater technical risk.     

High power pulsed lasers

Pulsed high power lasers can deliver sufficient energy on inertial fusion time scales (0.1–10 ns) to heat and compress DT fuel to fusion reaction conditions. Several laser systems have been examined for application to the fusion problem. Examples are Ndglass, CO₂, KrF, and I₂, etc. A great deal of developmental effort has been applied to the Ndglass laser and the CO₂ gas laser systems. These systems now deliver >10⁴ kJ and >20×10¹² W to inertial fusion targets. The Nova Ndglass laser is being constructed to provide >200 kJ and >200×10¹² W of 1 m radiation for fusion experimentation in the mid-1980s. For inertial fusion target gain, >100 times the laser input, it is expected that the laser must deliver 3–5 MJ of energy on the 10–20 ns time scale. This paper reviews the developments in laser technology and outlines approaches to construction of a 3–5 MJ driver.     

New cold nuclear fusion theory and experimental tests

A theory of neutron-induced tritium-deuterium fusion at room temperature is developed, based entirely on previously measured cross-sections of known nuclear reactions. The fusion process involves self-sustaining chain reactions: (1)n+⁶Li ⁴He+T and/orn+⁷Li⁴He+T+n, and (2) T+D ⁴He+n, in Li-D plasma or pellet surrounded by Li and other blankets and by neutron reflectors. The recent results of cold deuterium fusion reported by Fleischmann, Pons, and Hawkins are described in terms of this fusion process. Experimental evidence and tests of the chain reaction hypothesis are described.     

Thermodynamics Properties of Molten Salt Technology Assessment for New Generation Fusion Reactors

In this study, some important thermodynamic properties of the fusion reactor have been analyzed. The physical and chemical properties of molten salts have been extensively studied in the nuclear fusion program. In recent years, molten salts technology began to be used in some engineering areas, in the advanced nuclear field and especially in nuclear fusion reactor systems. Nowadays, Aries team has developed advanced designs by using the molten salts technology in order to get high thermodynamic and structural advantage on nuclear technology areas (Tillack et al. in Fusion Energ Des 65:215–261, 2003; Tillack et al. in Fusion Energ Des 49–50:689–695, 2000; El-Guebaly et al. in Fusion Energ Des 65:263–284, 2003). The Aries-St reactors are a 1000 MW fusion reactor system that based on a low aspect ratio ST plasma (Tillack et al. in Fusion Energy Des 65:215–261, 2003; Tillack et al. in Fusion Energy Des 49–50:689–695, 2000). The Aries team studies especially on liquid walls concepts and this liquid are used to increase neutronic performance of various structures of Aries-St reactors. In this research, candidate molten salts have been studied neutron effects on reactor performance which are the first wall (FW) and blanket. There are various candidate liquids that meet all the criteria such as Li₁₇Pb₈₃, flibe(Li₂BeF₄) and LiNaBeF₄, LiSn that are able to breed enough tritium. In this research, we used Li₁₇Pb₈₃, pure lithium and flibe as candidates that are in the Aries design. Montecarlo n-particle 4b-code is used for neutronics analysis and thermodynamic features. The value of tritium breeding ratio of the Aries-St reactors must be (TBR ≥ 1.1). This can be achieved in the region of LiPb/FW blanket of reactors. Aries-St spherical reactor has high heat flux (0.8 MW/m²) and NWL (6–8 MW/m²) in this region.     

High gain requirements and high field tokamak experiments

Operation at sufficiently high gain (ratio of fusion power to external heating power) is a fundamental requirement for tokamak power reactors. For typical reactor concepts, the gain is greater than 25. Self-heating from alpha particles in deuterium-tritium plasmas can greatly reducen/temperature requirements for high gain. A range of high gain operating conditions is possible with different values of alpha-particle efficiency (fraction of alpha-particle power that actually heats the plasma) and with different ratios of self heating to external heating. At one extreme, there is ignited operation, where all of the required plasma heating is provided by alpha particles and the alpha-particle efficiency is 100%. At the other extreme, there is the case of no heating contribution from alpha particles.n/temperature requirements for high gain are determined as a function of alpha-particle heating efficiency. Possibilities for high gain experiments in deuterium-tritium, deuterium, and hydrogen plasmas are discussed.     

The spherical pinch

This is essentially a review article covering several years of work on the spherical pinch (SP) concept of plasma formation and containment. Central to this concept is the creation of a hot plasma in the center of a sphere, plasma which is then compressed by strong imploding shock waves launched from the periphery of the vessel. The experimental program, which started with the classical cylindrical theta-pinch and continued with the inductive spherical pinch, has taken a turn, in recent times, with the discovery of the scaling laws governing spherical pinch experiments, which prescribe that high gas pressures are required for achieving fusion breakeven conditions. As a consequence, energy deposition in present spherical pinch devices is done through resistive, rather than inductive, discharges. In a pilot experimental program of modest initial condenser bank energy ( 1 KJ), we find that the instantaneous energy deposition in the central plasma can lead to temperatures of the order of 2 KeV, in agreement with the prediction of the Braginskii resistivity for such a plasma, and with the relation to the velocity of the diverging shock wave generated by the sudden deposition of energy into this plasma. Moreover, when the imploding shock waves contain the central plasma, we find the containment time to be as long as 5.4 sec and the plasma to be stable. In discharges in deuterium, neutrons are emitted close to 10⁷ per shot. From the experimental parameters of the plasma, one can derive a particle density for the shocked gas equal to 3.21×10¹⁹ cm^–3, a plasma temperature equal to 730 eV and a productn=1.73 × 10¹⁴ cm^–3· sec.Brief parts of this work are abstracted from previous works of the same author: C.A.S.I. (Canadian Aeronautics and Space Institute)Transaction,2, 21 (1969);Can. J. Phys.,58, 983 (1980);J. Fusion Energy,3, 199 (1983).     

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.     

Constraints on very-high-β, edge-stabilized tandem-mirror reactors

There are several tandem-mirror schemes which propose a very high and edge stabilization for the center-cell plasma ( being the ratio of the plasma pressure to the vacuum magnetic-field pressure). While the exact criteria for the edge stabilization are uncertain, it is possible to analyze the option space in which a very-high- mirror reactor would operate. The primary physics constraints on such a reactor are the energy balance at ignition, the buildup of He⁴ ash and the hot-particle( ^hot ), and the need for adiabatic conservation of the hot-particle gyro-orbits in the axial field gradients at the center-cell ends. There are also engineering constraints on the allowable wall loading and plant size. In this paper, a wall-stabilized tandem-mirror reactor is analyzed and is found to be an attractive device requiring low center-cell vacuum fields (of the order of 2 to 3 tesla). A primary requirement is that the plasma edge have a thermal conductivity near classical values.