Recent results of fusion triple product on EAST tokamak

High fusion triple product has been obtained in the advanced scenarios with high normalized beta (β_N) on the Experimental Advanced Superconducting Tokamak (EAST).A record value of n_(i0)T_(i0)τ_E~1.0?×?10~(19)m~(-3)ke V s for EAST deuterium plasma has been achieved,which is due to the formation of strong and broad internal transport barriers (ITBs) in n_e,T_eand T_iprofiles.Analysis shows that the strong ITB formation could be attributed to the reduction of transport from ITG modes.Based on the analysis,the physical mechanisms and methods to furtherimprove the plasma performance are discussed.     

Tritium breeding analysis of a tokamak magnetic fusion production reactor

With three-dimensional modeling and neutron transport analysis, a tokamak with a low technology blanket containing beryllium was found to have a tritium breeding ratio of 1.54 tritons per DT neutron. Such a device would have a net tritium production capability of 9.1 kg/yr from 450 MW of fusion power at 70% capacity factor.This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract W-7405-Eng-48.     

Method of emergency stoppage of nuclear fusion in a tokamak

Translated from Atomnaya Énergiya, Vol. 68, No. 2, pp. 116–118, February, 1990.     

The effect of wall loading limitations and choice of beta on the feasibility of advanced fuel fusion reactors

This paper investigates the effect of wall loading limitations and choice of plasma stability index beta on the feasibility of advanced fuel fusion reactors. Two new conceptual tools are introduced to facilitate this analysis: the “effective reactivity,” which includes all of the reaction-relevant parameters that determine the fusion power density, and the “critical radius,” which is the maximum allowable minor radius of a fusion reactor, beyond which the power generated in the plasma will exceed allowable loadings of radiant energy or neutrons on the first wall. It is shown that if high beta (greater than 0.2) fusion reactors are feasible, the high reactivity of the DT reaction cannot be fully exploited because of wall loading limitations. In addition, some high beta reactors with advanced fuels are also found to be wall loading limited, and to have excess reactivity, which can be traded off for lower magnetic fields, longer particle containment times, etc. Under certain circumstances, the reduced materials problems associated with some advanced fuels may outweigh the reactivity advantage of the DT reaction, and make one of them the reaction of choice for high beta fusion reactors.     

Transmutation characteristics of minor actinides in a low aspect ratio tokamak fusion reactor

The transmutation characteristics of minor actinides in the transmutation reactor of a low aspect ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculations coupled with a tokamak systems analysis were performed to determine optimal system parameters. The dependence of the transmutation characteristics, including the neutron multiplication factor, produced power, and the transmutation rate, on the aspect ratio A in the range of 1.5–2.0 was examined. By adding Pu239 to the transmutation blanket as a neutron multiplication material, it was shown that a single transmutation reactor producing a fusion power of 150 MW_th can destroy minor actinides contained in the spent fuels for more than 38 units of 1 GW_e pressurized water reactors (PWRs) while producing a power in the range of 1.8–6.8 GW_th.     

Albedo calculations for candidate fusion reactor materials used in the inboard side of a compact tokamak reactor and the effects of using such materials on the tritium breeding

In this paper the total neutron albedo and associated energy distributions for 10 candidate fusion reactor materials have been calculated. The angular distributions of reflected neutrons for monodirectional 14.1 MeV neutrons incident on slabs of Pb, Be, and W are presented and the dependence of albedo on neutron energy and incident angle has been investigated. Finally, the impact on the tritium breeding of the outboard blanket of the choice of material used in the inboard side of the reactor has been assessed. It has been found that the largest total neutron albedos are those of neutron multiplying materials, whilst among non-multiplying materials tungsten yields the largest albedo and B₁₀H₁₄ yields the lowest. Tritium breeding ratio (TBR) calculations have shown the inadequacy of the neutron albedo concept in predicting the impact of inboard materials on the TBR of the reactor.     

Overview of machine learning applications in fusion plasma experiments on J-TEXT tokamak

Machine learning research and applications in fusion plasma experiments are one of the main subjects on J-TEXT. Since 2013, various kinds of traditional machine learning, as well as deep learning methods have been applied to fusion plasma experiments. Further applications in the real-time experimental environment have proved the feasibility and effectiveness of the methods. For disruption prediction, we started by predicting disruptions of limited classes with a short warning time that could not meet the requirements of the mitigation system. After years of study, nowadays disruption prediction methods on J-TEXT are able to predict all kinds of disruptions with a high success rate and long enough warning time. Furthermore, cross-device disruption prediction methods have obtained promising results. Interpretable analysis of the models are studied. For diagnostics data processing, efforts have been made to reduce manual work in processing and to increase the robustness of the diagnostic system. Models based on both traditional machine learning and deep learning have been applied to real-time experimental environments. The models have been cooperating with the plasma control system and other systems, to make joint decisions to further support the experiments.     

Influence of the safety factor profile on the particle and heat transport in the Globus-M spherical tokamak

The advanced tokamak scenario is a promising operation scenario for ITER and fusion neutron sources. In this scenario the minimum value of the safety factor in the center of the plasma exceeds unity. In the compact spherical tokamak Globus-M, the formation of such conditions is possible with neutral beam injection at the current ramp-up phase. Due to the slower diffusion of current inside the plasma, a zone is formed with reduced heat and particle transport across the magnetic field, which affects the temperature and density profiles of the plasma. This leads to the peaked density profile formation and improvement of the energy confinement time. To achieve a high fraction of the bootstrap current, it is necessary to increase the plasma pressure. At the same time, the maximum allowable pressure is limited to the normalized beta limit.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

Effect of resonant magnetic perturbation on boundary plasma turbulence and transport on J-TEXT tokamak

The effect of resonant magnetic perturbation(RMP) on boundary turbulence and transport in J-TEXT plasma is experimentally investigated.Edge plasma fluctuations in discharges with and without the(m/n=3/1) RMP currents are diagnosed by using Langmuir probe arrays.It was found that fluctuations in the edge and scrape-off layer(SOL) regions decrease with the application of a 6 kA RMP.The broadband turbulence at the radial location of ρ～0.9 which has a characteristic frequency of 40-150 kHz was strongly suppressed when applying RMP,as was the radial turbulent particle flux and blob transport in the near-SOL region.These experimental findings make RMP a promising method of suppressing and controlling turbulence and particle transport in a plasma boundary.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components.To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×10¹⁵ n/(m²·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 10¹⁷ n/(m²·s) for a 1 MW/m² 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.     

Positron lifetime calculations of defects in fusion irradiated beryllium

Numerical quantum-mechanical positron lifetime calculations were performed for mono-vacancies, di-vacancies, tri-vacancies and small nano-voids containing helium and hydrogen in neutron irradiated beryllium. Helium and hydrogen atoms in the sample after the irradiation are considered as atoms forming interstitial O-type loops. Spherical clusters of vacancies are included in the calculations as a reference. It was found that the presence of He and H atoms significantly changes the positron lifetime in irradiated beryllium. A correlation between the positron lifetime and mutual position of vacancies in nano-voids and interstitial loops was established.     

Numerical prediction of pellet injection effects on core plasma fueling in the KSTAR tokamak

Since pellet injection into tokamak plasmas has been found to be an effective method for fueling and profile modification of core plasmas in tokamak experiments, a hypothetical injection of deutrium pellets into the KSTAR tokamak is numerically simulated in this work to investigate its influences on the fueling and transport of the core plasma depending on pellet parameters. A neutral gas shielding model and a pellet drift displacement model are used to describe the ablation and mass deposition from pellets on core plasma profiles. These models are coupled with a 1.5-dimensional (1.5D) core transport code to calculate the plasma density and temperature profiles responding to pellets injected into the target plasma. The simulation results indicate that a HFS (high field side) injection achieves more effective fueling due to a deeper pellet penetration into the core plasma, compared with a LFS (low field side) injection. The plasma density is found to increase during sequential pellet injections from both HFS and LFS, but the HFS case shows better fueling performance owing to a drift of the pellet ablatant in the major radius direction resulting in the deeper pellet penetration. Increasing the size and injection velocity of the pellet contributes to enhance the fueling efficiency. However, raising the power of neutral beam injection heating reduces the fueling efficiency because the pellet mass deposition is shifted toward the edge region in high temperature plasmas. It is concluded that the pellet size and injection direction among pellet and plasma parameters have the most dominant effects on fueling performance while the pellet velocity and heating power have relatively small influences on fueling.     

Optimization of nonthermal fusion power consistent with channeling of charged fusion product energy

If the energy of charged fusion products can be diverted directly to fuel ions, non-Maxwellian fuel ion distributions and temperature differences between species will result. To determine the importance of these nonthermal effects, the fusion power density is optimized at constant- for nonthermal distributions that are self-consistently maintained by channeling of energy from charged fusion products. For D-T and D-³He reactors, with 75% of charged fusion product power diverted to fuel ions, temperature differences between electrons and ions increase the reactivity by 40–70%, while non-Maxwellian fuel ion distributions and temperature differences between ionic species increase the reactivity by an additional 3–15%.     

Effects of resonant magnetic perturbations on radial electric fields in DIII-D tokamak

Gyrokinetic simulations of DIII-D tokamak equilibrium find that resonant magnetic perturbation (RMP) drives a neoclassical non-ambipolar electron particle flux, which causes a rapid change of equilibrium radial electric fields consistent with experimental observations during the suppression of the edge localized mode (ELM). The simulation results provide a support for the conjecture that RMP-induced changes of radial electric fields lead to the enhanced turbulent transport at the pedestal top during the ELM suppression (Taimourzadeh et al 2019 Nucl. Fusion 59 046005). Furthermore, gyrokinetic simulations of collisionless damping of zonal flows show that resonant responses to the RMP decrease the residual level of the zonal flows and damp the geodesic acoustic mode.