High Fusion Gain Research Articles

Head-on collision of compressed plasma jets shows promise for high-gain laser fusion

Creating high-density plasmas for the double cone ignition scheme

Observation of the colliding process of plasma jets in the double-cone ignition scheme using an x-ray streak camera

The double-cone ignition scheme is a novel approach with the potential to achieve a high gain fusion with a relatively smaller drive laser energy. To optimize the colliding process of the plasma jets formed by the CHCl/CD shells embedded in the gold cones, an x-ray streak camera was used to capture the spontaneous x-ray emission from the CHCl and CD plasma jets. High-density plasma jets with a velocity of 220 ± 25 km/s are observed to collide and stagnate, forming an isochoric plasma with sharp ends. During the head-on colliding process, the self-emission intensity nonlinearly increases because of the rapid increase in the density and temperature of the plasma jets. The CD colliding plasma exhibited stronger self-emission due to its faster implosion process. These experimental findings effectively agree with the two-dimensional fluid simulations.

Optimization of target compression for high-gain fast ignition via machine learning

The hydrodynamic scaling relations are of great importance for the design and optimization of target compression in laser-driven fusion. In this paper, we propose an artificially intelligent method to construct the scaling relations of the implosion velocity and areal density for direct-drive fast ignition by combining one-dimensional hydrodynamic simulations and machine learning methods. It is found that a large fuel mass and a high areal density required for high-gain fusion can be obtained simultaneously by optimizing the implosion velocity with less compression laser energy, taking full advantage of the separation of the compression and ignition processes in the fast ignition scheme. The obtained scaling relations are applied to the implosion design for the double-cone ignition scheme [Zhang et al., “Double-cone ignition scheme for inertial confinement fusion,” Philos. Trans. R. Soc., A 378(2184), 20200015 (2020)]. An optimized implosion is proposed with an areal density of 1.30 g/cm2 and a fuel mass of 215.7 μg with a compression laser energy of 168 kJ. Two-dimensional hydrodynamic simulations are further employed to validate the results. Our methods and results may be useful for the optimization of fusion experiments toward high-gain fusion.

Stopping power of high-density alpha-particle clusters in partially degenerated deuterium–tritium fuels

The state of burning plasma had been achieved in inertial confinement fusion (ICF), which was regarded as a great milestone for high-gain laser fusion energy. In the burning plasma, alpha particles incident on the cryogenic (warm dense) fuels cannot be simply regarded as single particles, and the new physics brought about by the density effects of alpha particles should be considered. In this work, the collective interaction between them has been considered, namely, the effect of the superposition of wake waves. The stopping power of alpha-particle clusters, i.e., the rate of energy loss per unit distance traveled has been calculated using both analytical and simulation approaches. In theory, we obtain the stopping power of alpha clusters in cryogenic (warm dense) fuel by the dielectric function method, which illustrates the importance of the effective interaction between particles. Simulation results using the LAPINS code show that the collective stopping power of the alpha cluster is indeed increased via coherent superposition of excitation fields (the excitation of high-amplitude wake waves). However, the comparison between simulation and theoretical results also illustrates a coherent–decoherent transition of the stopping power of the cluster. The initial conditions with various sizes and densities of the alpha clusters have been considered to verify the condition of decoherence transition. Our work provides a theoretical description of the transport properties of high-density alpha particles in warm dense cryogenic fuel and might give some theoretical guidance for the design of actual fusion processes.

Core-Pedestal Plasma Configurations in Advanced Tokamaks

Several configurations for the core and pedestal plasma are examined for a predefined tokamak design by implementing multiple heating/current drive (H/CD) sources to achieve an optimum configuration of high fusion power in a noninductive operation while maintaining an ideally magnetohydrodynamic (MHD) stable core plasma using the IPS-FASTRAN framework. IPS-FASTRAN is a component-based lightweight coupled simulation framework that is used to simulate magnetically confined plasma by integrating a set of high-fidelity codes to construct the plasma equilibrium (EFIT, TOQ, and CHEASE), calculate the turbulent heat and particle transport fluxes (TGLF), model various H/CD systems (TORIC, TORAY, GENRAY, and NUBEAM), model the pedestal pressure and width (EPED), and estimate the ideal MHD stability (DCON). The TGLF core transport model and EPED pedestal model are used to self-consistently predict plasma profiles consistent with ideal MHD stability and H/CD (and bootstrap) current sources. In order to evaluate the achievable and sustainable plasma beta, varying configurations are produced ranging from the no-wall stability to with-wall stability regimes, simultaneously subject to the self-consistent TGLF, EPED, and H/CD source profile predictions that optimize configuration performance. The pedestal density, plasma current, and total injected power are scanned to explore their impact on the target plasma configuration, fusion power, and confinement quality. A set of fully noninductive scenarios are achieved by employing ion-cyclotron, neutral beam injection, helicon, and lower-hybrid H/CDs to provide a broad profile for the total current drive in the core region for a predefined tokamak design. These noninductive scenarios are characterized by high fusion gain (Q ~ 4) and power (Pfus ~ 600 MW), optimum confinement quality (H98 ~ 1.1), and high bootstrap current fraction (fBS ~ 0.7) for Greenwald fraction below unity. The broad current profile configurations identified are stable to low-n kink modes either because the normalized pressure is below the no-wall limit or a wall is present.

Equation of state of tungsten-doped carbon based on QEOS model for laser fusion

Tungsten-doped diamond is employed as a promising ablator material in high-gain laser fusion target design. Unlike for pure carbon, reports on the equation of state (EOS) of tungsten-doped carbon are limited, particularly in the high-pressure range over Mbar, which is relevant to laser fusion. To complement the radiation-hydrodynamic simulations of laser fusion, we developed the EOS of tungsten-doped carbon by combining the quotidian EOS model and large-scale atomistic simulations, which provide fundamental material parameters. In this manner, the EOS of doped carbon can be efficiently constructed. The influence of tungsten doping on the diamond material parameters and EOS was observed. The application of the developed EOS was shown via typical radiation-hydrodynamic simulations of laser fusion.

High-βp scenario realized by the integration of internal and external transport barriers in the HL-2A tokamak

High-βp scenario addresses needs for an attractive tokamak fusion reactor design, where βp is the normalized plasma poloidal beta. High-βp experiments have been performed during recent years in the HL-2A tokamak. The high-performance region is realized by the integration of internal and external transport barriers which are dubbed as ITBs and ETBs, namely, double transport barriers (DTBs), with the high-power NBI heating. Generally, the ITB forms and becomes strong after the NBI injection on HL-2A. Subsequently, the edge ion temperature and toroidal rotation increase, as a result that the ETB or pedestal creates and L-H transition occurs. A high-βp (βp∼3) scenario is obtained with Ip∼110 kA and q95∼5, and another high-βp (βp∼2.7) scenario is also realized with Ip∼160 kA and q95∼4.2. The two high-βp scenarios are both characterized by a large-radius ITB and an ETB. However, the two scenarios are different. For the former there are strong large-radius ITB and weak ETB on radial ion temperature profiles, but for the later there are weak large-radius ITB and strong ETB. Although two high-βp scenarios on HL-2A have been accessed, however they are both transient. In addition, MHD instabilities in high-βp scenarios are also present for a high fusion gain and reactor relevant high fusion-density plasma.

Recent Progress in Modeling of CFETR Plasma Profiles from Core to Edge

Simulation studies of core plasmas and edge plasmas are performed in ASIPP to accommodate plasma profiles from core to edge based on physics modeling for the CFETR physics design. The coupling between the modeling for core plasmas and edge plasmas are described. The recent progress of the integrated core-pedestal simulations for core plasmas and the SOLPS simulations for plasmas in pedestal-SOL-divertor region are presented, respectively. In the optimization study for core plasma performance, moderate Z eff (~ 2.2) at pedestal top and a high ratio of density at separatrix to density at pedestal (n sep /n ped ~ 0.5) are found to be beneficial for the achievement of a high fusion gain in CFETR hybrid scenarios. The edge plasma simulations using the SOLPS code show that using either neon or argon impurity seeding can reduce the stationary power density on the divertor target to an acceptable level (< 10 MW/m2) but Z eff at pedestal top is still high (Z eff ~ 3). Future works to improve the consistence between the simulations of core and edge plasmas are discussed.

Constructing a new predictive scaling formula for ITER's divertor heat-load width informed by a simulation-anchored machine learning

Understanding and predicting divertor heat-load width λq is a critically important problem for an easier and more robust operation of ITER with high fusion gain. Previous predictive simulation data for λq using the extreme-scale edge gyrokinetic code XGC1 [S. Ku et al., Phys. Plasmas 25, 056107 (2018)] in the electrostatic limit under attached divertor plasma conditions in three major US tokamaks [C. S. Chang et al., Nucl. Fusion 57, 116023 (2017)] reproduced the Eich and Goldston attached-divertor formula results [formula #14 in T. Eich et al., Nucl. Fusion 53, 093031 (2013) and R. J. Goldston, Nucl. Fusion 52, 013009 (2012)] and furthermore predicted over six times wider λq than the maximal Eich and Goldston formula predictions on a full-power (Q = 10) scenario ITER plasma. After adding data from further predictive simulations on a highest current JET and highest-current Alcator C-Mod, a machine learning program is used to identify a new scaling formula for λq as a simple modification to the Eich formula #14, which reproduces the Eich scaling formula for the present tokamaks and which embraces the wide λqXGC for the full-current Q = 10 ITER plasma. The new formula is then successfully tested on three more ITER plasmas: two corresponding to long burning scenarios with Q = 5 and one at low plasma current to be explored in the initial phases of ITER operation. The new physics that gives rise to the wider λqXGC is identified to be the weakly collisional, trapped-electron-mode turbulence across the magnetic separatrix, which is known to be an efficient transporter of the electron heat and mass. Electromagnetic turbulence and high-collisionality effects on the new formula are the next study topics for XGC1.

A simplified economic model for inertial fusion

A simple model for the levelized cost of electricity (LCOE) of an inertial fusion power plant is developed. The model has 14 parameters. These have been designed to be technology agnostic, such that the model may be applied broadly to all variants of inertial fusion. It is also designed to allow easy use of proxies from existing technology. The variables related most intimately to the physics challenges of inertial fusion, such as gain and target cost, are treated as parameters such that requirements can be found without bringing complex physics into the model. A Monte Carlo approach is taken to explore the parameter space. The most important conclusion is that a combination of high gain (greater than 500) and high fusion energy yield per shot (greater than 5 GJ) together appear to unlock more cost competitive designs than those in the existing literature. Designs with LCOE as low as $25/MWh are found with optimistic but not obviously unrealistic inputs.This article is part of a discussion meeting issue ‘Prospects for high gain inertial fusion energy (part 1)'.

Direct drive with the argon fluoride laser as a path to high fusion gain with sub-megajoule laser energy.

Argon fluoride (ArF) is currently the shortest wavelength laser that can credibly scale to the energy and power required for high gain inertial fusion. ArF's deep ultraviolet light and capability to provide much wider bandwidth than other contemporary inertial confinement fusion (ICF) laser drivers would drastically improve the laser target coupling efficiency and enable substantially higher pressures to drive an implosion. Our radiation hydrodynamics simulations indicate gains greater than 100 are feasible with a sub-megajoule ArF driver. Our laser kinetics simulations indicate that the electron beam-pumped ArF laser can have intrinsic efficiencies of more than 16%, versus about 12% for the next most efficient krypton fluoride excimer laser. We expect at least 10% ‘wall plug' efficiency for delivering ArF light to target should be achievable using solid-state pulsed power and efficient electron beam transport to the laser gas that was demonstrated with the U.S. Naval Research Laboratory's Electra facility. These advantages could enable the development of modest size and lower cost fusion power plant modules. This would drastically change the present view on inertial fusion energy as being too expensive and the power plant size too large.This article is part of a discussion meeting issue ‘Prospects for high gain inertial fusion energy (part 1)'.

Projections of H-mode access and edge pedestal in the SPARC tokamak

In order to inform core performance projections and divertor design, the baseline SPARC tokamak plasma discharge is evaluated for its expected H-mode access, pedestal pressure and edge-localized mode (ELM) characteristics. A clear window for H-mode access is predicted for full field DT plasmas, with the available 25 MW of design auxiliary power. Additional alpha heating is likely needed for H-mode sustainment. Pressure pedestal predictions in the developed H-mode are surveyed using the EPED model. The projected SPARC pedestal would be limited dominantly by peeling modes and may achieve pressures in excess of 0.3 MPa at a density of approximately 3 × 1020m−3. High pedestal pressure is partially enabled by strong equilibrium shaping, which has been increased as part of recent design iterations. Edge-localized modes (ELMs) with &gt;1 MJ of energy are projected, and approaches for reducing the ELM size, and thus the peak energy fluence to divertor surfaces, are under consideration. The high pedestal predicted for SPARC provides ample margin to satisfy its high fusion gain (Q) mission, so that even if ELM mitigation techniques result in a 2× reduction of the pedestal pressure,Q&gt; 2 is still predicted.

Enhanced helium exhaust during edge-localized mode suppression by resonant magnetic perturbations at DIII-D

Effective helium confinement time, τp*,He, and its ratio with energy confinement time, τp*,He/τE, are key metrics quantifying the suitability of fusion plasmas for a continuous burn. Comparisons in the DIII-D tokamak of discharges with suppression of edge localized modes (ELMs) by resonant magnetic field perturbations (RMPs) to the corresponding unperturbed ELMy discharges found that both of these metrics were strongly reduced, by a factor of approximately 2, after application of the RMPs. This reduction in τp*,He during RMP ELM suppression was observed in the plasma core, edge, and pumping plenum, where higher neutral helium concentration during RMPs was also measured. These findings provide evidence that future devices employing RMP ELM suppression may meet or even exceed the helium exhaust provided by the ELMs themselves, reducing helium ash, and thus maintaining high fusion gain.

Fusion energy

This paper presents an overview of the main challenges that fusion research is facing on the road to a demonstration power plant. The focus is on magnetic confinement fusion. Most of the challenges are being addressed in the context of the ITER construction and exploitation. These include the demonstration of high fusion gain regimes of operation, the management of high heat and particle loads and the integration of the main technologies of a fusion power plant. In preparation of DEMO, reliable solutions for the breeding blanket and neutron resistant materials have to be developed.

The staged z-pinch as a potential high gain fusion energy source: An independent review, a negative conclusion

Previously published calculations predict that the “staged z-pinch” (SZP) can achieve 400 MJ of fusion yield on a Z-class machine. The SZP is touted to need no external preheat mechanism and no external pre-magnetization method. Instead, it is claimed that the imploding liner can adequately “shock preheat” the fuel and magnetic field diffusion through the liner can adequately magnetize the fuel. In this paper, we analyze a number of published SZP calculations and demonstrate that the calculations have major errors—the computer code used to do the calculations does not appear to be accurately solving the physical model it is intended to solve. A variety of independent analyses lead to this conclusion. This conclusion is confirmed by detailed one-dimensional magnetohydrodynamic (MHD) calculations conducted on different computer codes using a variety of proposed SZP operating parameters. Although using parameters similar or identical to the published calculations, our MHD calculations do not reach fusion conditions; there is no conceivable modification of the parameters that would lead to high-gain fusion conditions using these other codes. Our analyses and a review of the magnetized target parameter space leads to further conclusion that the SZP should not be considered to be a potential high-gain fusion source.

Radio-frequency current drive for thermonuclear fusion reactors

Principal research on energy from thermonuclear fusion uses Deuterium-Tritium plasmas magnetically trapped in toroidal devices. As major scientific problem for an economic (i.e., really feasible) reactor, we must understand how to lead strongly heated plasmas to sustain a high fusion gain while large fraction of current is self-produced via the presence of strong pressure gradient. To suppress turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary. However, tools envisaged so far in ITER (International Thermonuclear Experiment Rector) are unable accomplishing this task that requires efficiently and flexibly matching the natural current profiles of plasma. Consequently, viability of a thermonuclear reactor should be problematic. Multi-megawatt radio-frequency (RF) power coupled to plasma would produce the necessary CD, but modelling results based on previous understanding found difficult the extrapolation of this CD concept to reactor conditions of high temperature plasma, and greater flexibility of method would also be required. Here we present new model results based on standard quasilinear (QL) theory that allow establish conditions to drive efficiently and flexibly the RF-driven current at large radii of the plasma column, as necessary for the goal of a reactor.

Laser-driven strong magnetostatic fields with applications to charged beam transport and magnetized high energy-density physics

Powerful nanosecond laser-plasma processes are explored to generate discharge currents of a few 100 kA in coil targets, yielding magnetostatic fields (B-fields) in excess of 0.5 kT. The quasi-static currents are provided from hot electron ejection from the laser-irradiated surface. According to our model, which describes the evolution of the discharge current, the major control parameter is the laser irradiance Ilasλlas2. The space-time evolution of the B-fields is experimentally characterized by high-frequency bandwidth B-dot probes and proton-deflectometry measurements. The magnetic pulses, of ns-scale, are long enough to magnetize secondary targets through resistive diffusion. We applied it in experiments of laser-generated relativistic electron transport through solid dielectric targets, yielding an unprecedented 5-fold enhancement of the energy-density flux at 60 μm depth, compared to unmagnetized transport conditions. These studies pave the ground for magnetized high-energy density physics investigations, related to laser-generated secondary sources of radiation and/or high-energy particles and their transport, to high-gain fusion energy schemes, and to laboratory astrophysics.

Direct heating of a laser-imploded core using ultraintense laser LFEX

A CD shell was preimploded by two counter-propagating green beams from the GEKKO laser system GXII (based at the Institute of Laser Engineering, Osaka University), forming a dense core. The core was predominantly heated by energetic ions driven by the laser for fast-ignition-fusion experiment, an extremely energetic ultrashort pulse laser, that is illuminated perpendicularly to the GXII axis. Consequently, we observed the D(d, n)3 He-reacted neutrons (DD beam-fusion neutrons) at a yield of n sr. The beam-fusion neutrons verified that the ions directly collided with the core plasma. Whereas the hot electrons heated the whole core volume, the energetic ions deposited their energies locally in the core. As evidenced in the spectrum, the process simultaneously excited thermal neutrons with a yield of n sr, raising the local core temperature from 0.8 to 1.8 keV. The shell-implosion dynamics (including the beam fusion and thermal fusion initiated by fast deuterons and carbon ions) can be explained by the one-dimensional hydrocode STAR 1D. Meanwhile, the core heating due to resistive processes driven by hot electrons, and also the generation of fast ions were well-predicted by the two-dimensional collisional particle-in-cell code. Together with hot electrons, the ion contribution to fast ignition is indispensable for realizing high-gain fusion. By virtue of its core heating and ignition, the proposed scheme can potentially achieve high-gain fusion.

Impact of impurity seeding and divertor conditions on transitions, pedestal structure and ELMs

Future devices will require a high scrape-off later (SOL) density and impurity seeding to avoid high-Z sputtering. However, these operational parameters are not included in present-day scaling laws, making extrapolations to larger devices difficult. As such, understanding the physics of such effects is vital in order to design the operational scenarios most favorable to high fusion gain. This review presents the favorable lowering of L-H transition power by changing to metal walled devices and sumarises the effects currently thought to be responsible for how SOL geometry can play a role in determining this threshold. Experimental observations on changes to the pedestal structure with main ion fuelling and low-, medium-, and high-Z impurity seeding are presented. These results, from several devices, show that main ion fuelling or high density operation can result in a lower pedestal top pressure, and hence reduced stored energy, while impurity seeding can recover this lost pressure. Particular focus is given to nitrogen seeded discharges and the recovery of pedestal parameters (notably high ) in JET and AUG since the changeover to metal walls in these devices. Lithium seeding is also emerging as a strong actuator in pedestal dynamics, with results ranging from a prolonged inter-ELM period to completely ELM-free scenarios on different devices. ELM dynamics are also presented in each section, with nitrogen seeding offering a probe into the structure of the ELM and demonstrating the difference between the initial ELM crash, likely due to a sharp MHD event, and a prolonged second phase, the origin of which remains unkown. Finally, modelling of the pedestal in impurity seeded scenarios reveals a common effect in the position of the density profile. Either through mode excitation near to the separatrix or an altered fuelling profile, seeding of impurities results in an inward shift of the density profile. This inward shift improves MHD stability, allowing access to regions of higher pressure gradient and current density, and, hence, offering a promising operational scenario for future devices.

Evaluation of CFETR key parameters with different scenarios using system analysis code

This paper presents a system analysis on both phases of China Fusion Engineering Test Reactor (CFETR) under the same structure of toroidal field coil with a magnetic field on axis of 6–7T and device size of major radius of R∼6m. Phase I is designed with moderate gain aimed at exploring steady-state scenarios with DEMO relevant plasmas. One operating mode with higher magnetic field and conservative physics is presented, which could serve as a backup CFETR baseline in case of shortfall in physics performance. Phase II targets high gain with power plant relevant requirements and objectives, including more than 1GW of fusion power, high fusion gain for DEMO validation and more ambitious assumptions based on advanced physics. A system analysis code, General Atomics System Code (GASC) [R. Stambaugh, V. Chan, et al., Fusion nuclear science facility candidates, Fusion Sci. Technol., 59 (2001) 279–307] is used to evaluate key parameters of scenarios for both phases.