Simulations of NBI fast ion loss in the presence of toroidal field ripple on EAST

NBI fast ion losses in the presence of the toroidal field ripple on EAST have been investigated by using the orbit code GYCAVA and the NBI code TGCO. The ripple effect was included in the upgraded version of the GYCAVA code. It is found that loss regions of NBI fast ions are mainly on the low field side near the edge in the presence of ripple. For co-current NBIs, the synergy effect of ripple and Coulomb collision on fast ion losses is dominant, and fast trapped ions located on the low field side are easily lost. The ripple well loss and the ripple stochastic loss of fast ions have been identified from the heat loads of co-current NBI fast ions. The ripple stochastic loss and the collisioninduced loss are much larger than the ripple well loss. Heat loads of lost fast ions are mainly localized on the right side of the radio frequency wave antennas from the inside view toward the first wall. For counter-current NBIs, the first orbit loss due to the magnetic drift is the dominant loss channel. In addition, fast ion loss fraction with ripple and collision for each NBI linearly increases with the effective charge number, which is related to the pitch angle scattering effect.     

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions on EAST

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions in the EAST tokamak have been studied in detail by using the orbit-following code GYCAVA and the NBI code TGCO. Beam ion losses with the wall boundary are smaller than those with the last closed flux surface boundary. In contrast to heat loads on the wall without radio frequency wave(RFW)antennas, heat loads on the wall with RFW antennas are distributed more locally near the RFW antennas. The direction of the toroidal magnetic field dramatically affects the final positions of lost fast ions, which is related to the magnetic drift. The numerical results on heat loads of beam ions corresponding to different toroidal magnetic fields are qualitatively consistent with the experimental results. Beam ion losses increase with the beam energy for the co-current NBIs and the counter-perpendicular NBI. We have studied the behavior of fast ions produced by a small section neutral beam(beamlet) by using the numerical tool NBIT. The distributions of the loss fraction of beamlet fast ions peaked near the edge of the beam section for the counter-current NBIs, and they are related to the injection angle. This indicates that the first-orbit losses can be reduced by changing the shape of beam cross section.     

Symplectic Simulations of Radial Diffusion of Fast Alpha Particles in the Presence of Low-Frequency Modes in Rippled Tokamaks

The symplectic Hamiltonian guiding centre code which enables efficient calculation of charged particle trajectories and diffusion coefficients has been applied to fast ion motion in magnetically perturbed tokamak plasmas. Particularly fusion born alpha particle drift motion, in constant of motion space, is examined in the presence of low mode-number neoclassical tearing mode (NTM) perturbation in a toroidally rippled tokamak. The main focus of this study is to investigate the dependence of the radial diffusion coefficient of energetic ions on the perturbation strength and on the localization of the perturbation. The resonance between bounce motion and toroidal field ripples plays a significant role in this context. The presence of NTMs results in substantial enhancement of radial diffusion coefficient for passing particles. Depending on the strength and localization of the NTM it can cause enhancement or degradation of the radial ripple diffusion coefficient of trapped particles.     

Stability analysis of Alfvén eigenmodes in China Fusion Engineering Test Reactor fully non-inductive and hybrid mode scenarios

In this paper, NOVA/NOVA-K codes are used to investigate the stability of Alfvén eigenmodes(AEs) in the China Fusion Engineering Test Reactor(CFETR). Firstly, the stability of AEs excited by energetic alpha particles is investigated. For the fully non-inductive scenario, it is found that all AEs are stable, and the least stable toroidal mode number is n= 8. However, for the hybrid mode scenario, it is found that many AEs are unstable, and the least stable toroidal mode numbers are n= 7, 8. Secondly, the effect of energetic alpha-particle parameters and beam ions on AE stability is also presented. The threshold of the least stable AE is about β_(crit,α) = 1.12%,crit,less than the value of alpha-particle beta(β_α=1.34%). The result demonstrates that the AEs excited by alpha particles are weakly unstable. The effect of the beam ions on AE stability is found to be very weak in CFETR.     

Progress in CFETR Power Stations Concept Study

While EAST experiment was running in 2012, the project of the China fusion engineering test reactor (CFETR) concept design was started. This ITER-like tokamak system will be the second full superconducting tokamak in China based on EAST technology. In phase I, it has 50–200 MW heat output for demonstrating power generation. The fusion power stations contain complete structure of fusion power plant (FPP) which do not appear in the ITER and huge HV substation which receives power from the 500 kV transmission grid for powering its pulsed power electric network (PPEN) and steady-state electric power network. Furthermore, its structure of turbine generator of FPP is similar to that of a nuclear power station of the pressurized-water reactor. This paper describes the typical CFETR loads and put forward the requirements of short circuit capacity of HV grid. It analyzes different strategies of putting the generator power to the grid, i.e. on the 500 kV grid for future DEMO power structure or 66 kV medium-voltage local grid for self-use. In period between twice burning plasma, conceptual solutions are presented to maintain thermal circuit operation.     

The Effect of Externally Applied Resonant Magnetic Perturbations on Fusion Product Dynamics in Toroidal Plasmas: Numerical Simulation

The suppression of edge-localized modes by means of externally induced resonant magnetic perturbations (RMPs) has been investigated extensively on present-day tokamaks. In this paper we examine the modification of the loss of fusion born α-particles as effected by the application of RMPs in tokamak plasmas. This study was performed by means of test-particle simulations. To simplify the calculations we use a toroidal magnetic field model with circular magnetic flux surfaces. The transport properties of energetic α-particles are investigated during a 3 s time interval by tracing the test-particle ensemble with each particle trajectory following by integration of the full orbit equation. Three regimes of particle losses are identified during the evolution of the particle ensemble. A natural consequence of RMP excitation is the formation of magnetic islands together with stochastic magnetic layers at the plasma edge. The formation of these resonant magnetic field structures are associated with irregularities of the energetic α-particle orbits, which can substantially increase the loss of slowing down α-particles from the plasma periphery. At the same time the first orbit losses of fusion alphas are practically unaffected by RMPs.     

Effects of ferromagnetic components on energetic ion confinement in ITER

By using a fully three dimensional magnetic field orbit-following Monte-Carlo code, the energetic ion confinement was investigated for the current conceptual design of the ferromagnetic components in ITER which will be employed for reducing the toroidal magnetic field (TF) ripple. The ferromagnetic insert is effective in the reference standard scenario with Q = 10 (Scenario No. 2) and steady state scenario with Q = 5 (Scenario No. 4) to improve the energetic ion confinement. Over-compensation appears at half of the full toroidal magnetic field and its effect becomes stronger when the quantity of the ferromagnetic insert is increased in order to more reduce the TF ripple at the full toroidal magnetic field. Though the current design is acceptable, whether to increase the ferromagnetic insert to achieve lower TF ripple amplitude at the full field operation depends on how prospected are possibilities of lower field operations. Planned test blanket modules do not induce large loss (<1%) at the full field in Scenario No. 4. At the half field, however, the loss reaches ∼10% for the alpha particles due to localized large TF ripple.     

Finite Larmor radius effects on ripple diffusion in tokamaks

The confinement of alpha particle is important in tokamaks. The trajectory or confinement of alpha particles is often calculated with guiding center approximation. The features of spherical tokamak can break the approximation. We investigate finite Larmor radius effect. The orbit calculated by equation of motion is different, especially in toroidal direction, from that by guiding center equations. This difference causes difference of ripple resonance energy and the difference of the peaks of the diffusion coefficients. Furthermore, because of the additional free parameter, Larmor phase, the peaks of diffusion coefficient calculated by equation of motion are broader than those by guiding center equations.     

Numerical Study on Ripple Loss of Neutral-Beam-Injected Fast Ions in EAST

A systematic comparative study on the behaviors and loss processes of energetic beam ions in the rippled toroidal field on Experimental Advanced Superconducting Tokamak (EAST) is carried out by numerical simulations. The predicted loss fractions of either co-injected or counter-injected neutral beam ions for typical EAST experiments are 9–16 %; while for low current experiments, the ripple loss domain is enlarged with the increase of the safety factor q, resulting in enhanced beam ions loss. In addition, Counter-injected ions give rise to more lost fractions in the relatively high-energy range, and fewer of them distribute into the core plasma region, which suggest that the co-injection scheme is somewhat preferable for plasma heating. Moreover, the total losses of energetic beam ions in both co-injection and counter-injection geometries are seen to be due entirely to the delayed losses owing to a synergistic effect of collisions and ripple. Finally, potential first wall damage appears to be avoidable for long pulse neutral beam injection scenarios.     

Symplectic Simulation of Fast Alpha Particle Radial Transport in Tokamaks in the Presence of TF Ripples and a Neoclassical Tearing Mode

A Hamiltonian guiding centre drift orbit code based on a symplectic integration algorithm, which enables the efficient calculation of particle trajectories and diffusion coefficients, is applied to fast alpha particle motion in magnetically perturbed tokamak plasmas. In particular, fast ion drift motion is examined in the presence of a stationary, low mode-number MHD magnetic perturbation in a toroidally rippled tokamak with circular flux surface. The main focus of our study is to investigate the dependence of the radial diffusion coefficient of energetic ions on their energy, on the perturbation strength and the localization of the perturbation. As expected, the resonance between bounce motion and toroidal field ripples plays a significant role in this context. For an ensemble of fast ions uniformly distributed in toroidal angle but with a given poloidal starting position their radial transport coefficient takes on higher values in the neighbourhood of resonance speeds and can exhibit there local minima, i.e. it shows an M-shaped speed dependence around resonances for sufficiently strong ripple perturbations. Expectedly, the addition of a modelled low-mode number neoclassical tearing mode perturbation will modify the pure ripple resonance structure of the radial diffusion coefficient. Depending on the strength and localization of the MHD mode it can cause enhancement or degradation of the radial ripple diffusion coefficient.     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

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.     

Comparison of calculations of fusion product losses from a tokamak

A comparison is made of models for 3.5-MeV alpha particle loss to the first wall of an axisymmetric tokamak, including calculations of the loss fraction, wall loading profile, and sensitivity to machine parameters.     

Integrated Operating Scenario to Achieve 100-Second,High Electron Temperature Discharge on EAST

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape,divertor heat flux,particle exhaust,wall conditioning,impurity management,and the coupling of multiple heating and current drive power.A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA,q_(95)≈7.0,a peak electron temperature of 4.5 keV,and a central density n_e(0)~2.5×10~(19) m~(-3).The plasma current was nearly non-inductive(V_(loop) 0.05 V,poloidal beta ~0.9) driven by a combination of 0.6 MW lower hybrid wave at 2.45 GHz,1.4 MW lower hybrid wave at 4.6 GHz,0.5 MW electron cyclotron heating at 140 GHz,and 0.4 MW modulated neutral deuterium beam injected at 60 kV.This progress demonstrated strong synergy of electron cyclotron and lower hybrid electron heating,current drive,and energy confinement of stationary plasma on EAST.It further introduced an example of integrated "hybrid" operating scenario of interest to ITER and CFETR.     

高温颗粒落入水中瞬态过程的数值模拟

以液膜质量、动量和能量守恒方程为基础提出了一个高温颗粒在冷却剂中运动的瞬态理论模型.由于高温颗粒在冷却剂中运动的瞬态复杂性,利用适合于求解大条件数的Gear算法对该模型进行了数值模拟.求解过程中采用自适应技术处理高温颗粒在冷却剂中运动时周围蒸汽膜不断变动的动边界问题,得到高温颗粒下落速度的理论计算值,并与实验值进行比较,得到了蒸汽膜的温度、厚度以及作用力等瞬态数值模拟结果.     

Alfvén eigenmode stability analysis and energetic particle transport prediction for CFETR hybrid scenario

The hybrid scenario is a projection for CFETR operation with high plasma current and density. Therefore, the energetic particles (EPs) generated by fusion reactions can destabilize Alfvén eigenmodes (AEs), which could result in significant EPs loss and redistribution. Both the eigenvalue code NOVA-K and the wrapped local stability code TGLFEP are used to analyze AE stability. The simulation indicates the beta-induced Alfvén eigenmodes with n>5 in the core region are the most unstable. The NOVA-K code is used to benchmark the critical density gradient calculated by TGLFEP. In addition, the EPtran code is employed to predict EP transport induced by unstable AEs and turbulence, which reduce EP density in the core and drive approximately 30% EP transport from the core to the edge, thus the EP density profile flattens and EPs with lower energy deposit near the edge.     

Assessment of DEMO challenges in technology and physics

The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5 MW/m², leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.     

A new approach to plasma profile control in ITER

Advanced tokamak operation in ITER, such as the steady-state and hybrid modes, requires an active real-time feedback control of plasma profiles to achieve the advanced regimes for sustained operation. In this work, we have explored a potentially robust control technique that simplifies the active real-time control of electron temperature and safety factor profiles in ITER. As a new and simple approach, static responses of the plasma profiles to power changes of auxiliary heating and current drive are modelled and updated in real-time, differing from the techniques which use a dynamic model deduced from identification experiments, or even a simplified explicit model. To allow real-time update of the plasma profile response model, the underlying physics is simplified with several assumptions. The electron temperature profile response is modelled by simplifying the electron heat transport equation. The safety factor profile response is modelled by directly relating it to the changes of source current density profiles. The required actuator power changes are calculated using the singular value decomposition technique, taking the saturation of the actuator powers into account. The potential of this control technique has been tested by applying it to simulations of the ITER hybrid mode operation using CRONOS. In these simulations, the electron temperature and safety factor profiles were well controlled either independently or simultaneously.     

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.     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.