Search

The negative triangularity tokamak (NTT) is a unique reactor concept based on 'power-handling-first' philosophy with the heat exhaust problem as the leading concern. The present paper exposes a reactor concept using L-mode edge based on NTT configuration, providing merits of no (or very weak) edge-localized modes, larger particle flux and large major radius for power handling. It is shown that a reasonably compact (R-p from 9 m to 7 m) NTT reactor is possible by achieving higher confinement improvement (H-H = 1.5) and/or by utilizing reasonably higher magnetic field (B-max = 15.5 T). Current physics basis and critical issues on its scientific and technical feasibility are discussed.

Tungsten (W), a promising candidate as divertor plasma facing material in magnetic fusion devices, is anticipated to promptly redeposit when sputtered or evaporated from surface due to its small ionization energy and long gyro radius. Using an artificial factor for the reaction cross sections, effects of ionization lengths to the re-deposition rate was studied by a newly developed particle-in-cell code. Treating numbers of particles in a super particle, electric charge, and mass as particle variables in the code, a special scheme for ionization and recombination was developed and used for the calculation. Simulations on W test particles with imaginary properties (neglecting the electric force) revealed the effects of ejection angles. Simulations with secondary electrons from surface showed that the sheath potential is weakened and the re-deposition rate becomes small. It was found that the multi-ionization as well as the ionization mean-free-path influences the re-deposition rate in both simulations. (© 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

One-dimensional SOL-divertor plasma fluid simulation code which considers anisotropy of ion temperature has been developed so as to deal with sheath theory self-consistently. In our fluid modeling, explicit use of boundary condition for Mach number M at divertor plate, e.g., M = 1 , becomes unnecessary. In order to deal with the Bohm condition and the sheath heat transmission factors at divertor plate self-consistently, we introduced a virtual divertor (VD) model which sets an artificial region beyond divertor plates and artificial sinks for particle, momentum and energy there to model the effects of the sheath region in front of the divertor plate. Validity of our fluid model with VD model is confirmed by showing that simulation results agree well with those from a kinetic code regarding the Bohm condition, ion temperature anisotropy and supersonic flow. We also show that the strength of artificial sinks in VD region does not affect profiles in plasma region at least in the steady state and that sheath heat transmission factors can be adjusted to theoretical values by VD model. Validity of viscous flux is also investigated.

Divertor plasma detachment is the most promising candidate to reduce the divertor heat load in fusion reactors. Present understanding of detachment physics is not sufficient to adequately reproduce experimental observations. Understanding and control of detachment physics is indispensable to design the divertor in future machines. To improve the quality of divertor modeling and reveal limitations of the detachment physics built into state-of-the-art codes, an integrated divertor code SONIC has been applied to modeling of the JT-60U detached divertor plasma. In this study, the radial diffusion coefficient in the private region or the far SOL region is increased to investigate the influence of radial plasma transport on detachment characteristics. Saturation of the reduction in ion flux after roll-over is improved by the radial transport enhancement, while the radial profile at the mid-plane agreed with the experimental data.

Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor.

A 1D SOL–divertor plasma simulation code by introducing the anisotropic ion temperature with virtual divertor model has been developed. By introducing the anisotropic ion temperature directly, the second-order derivative parallel ion viscosity term in the momentum transport equation can be excluded and the boundary condition at the divertor plate will not be required in the simulation. In order to express the effects of the divertor plate and accompanying sheath implicitly, a virtual divertor model which has artificial sinks for the particle, momentum and energy has been introduced. Periodic boundary condition becomes available by the use of the virtual divertor model. By using this model, SOL–divertor plasmas which satisfy the Bohm condition has been successfully obtained. The dependence of the ion temperature anisotropy on the normalized mean free path of ion and the validity of the parallel ion viscous flux for the Braginskii expression and the limited one are also investigated.

Particle-in-cell (PIC) simulations are the very powerful method to understand physics of edge plasmas in fusion devices. Poisson solver with a regular grid in a rectangular vessel is very fast, but it cannot simply be applicable to simulations in complicated divertor configuration. In this paper, a simple and fast Poisson solver with arbitrary boundary shape and condition for PIC simulation is proposed. Boundary points are essential for this Poisson solver, on which the charge density is appropriately given and the boundary condition is satisfied. For particle simulations in the open system, it is necessary to distinguish whether a traced particle moves freely or hits the wall boundary during a time step. A new method for PIC simulation to discriminate a particle position before the boundary or beyond the boundary is also presented. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Simulations with an integrated code TOPICS-IB showed that a small pellet can significantly reduce the ELM energy loss by penetrating deeply into the pedestal and triggering high-n ballooning modes localized near the pedestal top, with conditions; the injection from the low-field-side with a speed fast enough to approach the pedestal top when the pedestal pressure is about 95% of natural ELM onset. The effectiveness of the above suitable conditions of pellet injection for ELM pacing has been confirmed by JT-60U and then ITER simulations. The pellet particle content required for ELM pacing is larger for the pedestal plasma with higher density and farther from the stability boundary of ideal ballooning mode near the pedestal top. For an ITER standard scenario, the required pellet particle content is about a few % of pedestal particle content, which gives the physics background to the present design value. Simulations also showed that fueling pellets can be injected from the high-field-side just after ELM pacing pellets without disturbing the pacing. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The simulation study of nonlocal transport from edge to core in tokamak plasmas is performed using the 4-field reduced MHD model. A toroidally-elongated cylindrical particle source is applied in the plasma edge, after saturation of the resistive ballooning turbulence is attained. After a short time, the source is switched off and plasma response is investigated in detail. The nonlocal transport appears at the location far from the edge source. It is found that the particle source induces (0,0) and (±1, 0) modes of density fluctuations, where (m, n) indicates the set of poloidal mode number m and the toroidal mode number n. These modes interact with each other by the nonlinear and/or toroidal couplings. The symmetry of (±1, 0) modes breaks after switching-off the source and the formation of the spiral structure with poloidal rotation is observed, which yields a connection between core and edge regions. In this simulation, the convective cell mode such as (1,0) mode contributes the nonlocal transport. The simulation result indicates that two dimensional transport plays an essential role to produce the nonlocal transport. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The Monte-Carlo (MC) approach has a lot of flexibility in impurity transport modeling in the SOL and divertor region. However, in the divertor plasma simulation with the noble impurity seeding, characteristic time of the impurity transport especially in the sub-divertor chamber is long because the MC calculation of the impurity gas transport can be finished only by exhaust. The impurity MC calculation for such long exhaust processes is difficult in a series of the iterative calculation of a suite of integrated divertor codes SONIC. In order to overcome such a problem, a backflow model has been developed. Amount of the backflow flux from the sub-divertor chamber to the divertor region is evaluated in advance, and then simulating impurity flux is injected from the exhaust slot to the divertor region like a backflow. By this model, the MC calculation time is reduced significantly and iterative calculation of SONIC becomes possible within a reasonable calculation time. As a demonstration, the SONIC code with the backflow model has been applied to investigation of power handling in JT-60SA divertor. The SONIC simulation showed that low divertor heat load (< 10 MW/m2) with the low SOL density (< 1.5 × 1019m–3), which is required in the full non-inductive current drive scenario, was achieved by the Ar gas puffing of 0.86 Pa m3/s. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

In ITER and future fusion devices, wall erosion caused by transient heat loads, edge-localized modes (ELMs) and disruptions is a serious concern. Suppression of the erosion occurs when a significant vapor emission happens under the heat loads, which is known as vapor shielding. Estimation of the suppressed erosion amount is essential for life time analysis and reactor design. In this study, we developed a particle-in-cell simulation code for the vapor shielding, PIXY, and applied the code to the erosion estimation for Be and W walls under a pulsed heat flux of 1–20 GW m−2 within 0.2 ms, which simulates a short-duration ELM pulse. Heat fluxes reaching the wall are recorded separately for electrons and ions, and reveal effective physical processes during the vapor shielding at Be and W walls. Then, for these walls, the total erosion amounts with and without vapor shielding are evaluated and compared. Estimated total erosion is larger for the Be wall up to 10 GW m−2. The Be wall erosion becomes saturated for the heat flux >10 GW m−2 due to its effective shielding, while the estimated W wall erosion eventually overcomes the Be wall erosion. In the short-time scale studied in this paper, effective radiation cooling by W vapor clouds is not observed. A simple wall life time analysis based on the estimated erosion rate shows 10 mm of W wall will survive 107 ELM loads with energy density less than 1 MJ m−2.

Edge localized mode (ELM) must be eliminated which enhances the erosion of divertor plates in the Hmode operation of tokamak reactors. Suppression of ELM has been experimentally achieved by the resonant magnetic perturbation (RMP) with multipartite coils. In a DEMO reactor with strong neutron flux, however, it is desired the coils near the first wall not to be put in. We propose an innovative concept of the RMP for tokamak DEMO reactors without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed magnetic field is naturally formed in the axisymmetric toroidal magnetic field through the helical ferritic steel inserts (FSIs). The perturbation amplitude in the plasma pedestal region can easily be set above several 10-4 of the toroidal field strength in the DEMO reactor condition, which is enough for the RMP to mitigate/suppress ELMs.

In SOL-divertor plasmas, a condition on the plasma flow velocity, called the Bohm criterion, is required for the formation of a stable sheath in front of a divertor plate. While, for a single-ion species, the usual analytical procedure gives the criterion as a unique solution for the incident ion flow velocity, for two-ion-species plasmas the procedure results in a relation between the ion flow velocities. We carried out full particle simulations with a one-dimensional particle code PARASOL, and clarified in detail the dependence of the condition for two-ion-species plasmas on collisionality and ions mass (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

We have developed a self-consistent integrated simulation model of core and scrape-off-layer (SOL)/divertor transport. This model enables the user to investigate operational scenarios with compatible high-confinement core plasmas and detached divertor plasmas. A 1.5D core code TOPICS-IB and a 2D divertor code SONIC are successfully coupled using a Multiple Program Multiple Data parallel computing system. This integrated code, which can include Monte-Carlo models for neutrals and impurities, provides efficient plasma/neutral simulations. The dynamic simulation for the L/H transition in JT-60SA is carried out by this integrated code using a current-diffusive ballooning mode (CDBM) transport model that includes the E×B shearing effect. The impact of SOL/divertor transport on the L/H transition is studied (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Power handling in the divertor is one of the most important issues for fusion tokamak reactor design. In this study, the effect of the divertor leg length on power handling in the SlimCS DEMO reactor is investigated for future divertor design. Simulation with the SONIC code suite shows enhancement of recycling in the outer divertor region due to effects of the long divertor leg and the deep V-shaped corner. As a result, the ion and electron temperature at the outer divertor significantly decreases. In addition, the peak of the Ar impurity radiation, which is located close to the outer divertor target in the standard divertor geometry, moves upstream. The former effect decreases the heat load to the outer divertor due to ion conduction and convection, while the latter reduces the heat load due to the impurity radiation. The peak of the total heat load on the outer divertor decreases by about 30 % (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

An accurate calculation of heat load on the divertor plates in a tokamak must take into account kinetic effects, present when the electron velocity distribution function (EVDF) in the plasma column departs from a thermal distribution. In this work PARASOL-1D, a particle-in-cell code with a Monte-Carlo binary collision model, is used to find and explain the electron heat flux-limiting factor αe and the heat transmission coefficient (HTC) γe in the complex SOL. We develop and test two simple models for these kinetic factors that take as input the temperature of the SOL and the temperature of the electrons striking the divertor plates. Both models assume a piecewise EVDF, with a symmetric bulk electron population and a high-energy tail of electrons moving only in the direction of the nearest diverter plate. The model EVDF are fit to the simulation-derived EVDF by allowing the variables for the densities and temperatures in the bulk and tail to vary independently. The block model, which assumes a bulk population of infinite parallel temperature in the bulk, is found to reproduce the kinetic factors for a wide range of conditions in the complex SOL much better than both the classical and Gaussian models (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

In order to improve the detachment modelling, effects of various processes on ion flux reduction other than the volume recombination are investigated by using a suite of integrated divertor codes SONIC. The increase in the radial diffusion loss at the private edge decreases the ion density and the ion flux by 15% at most. The supersonic flow effect on flux reduction is small because the supersonic flow occurs just in front of the target under the present simulation condition. Due to effects of the wall pumping or D neutral recycling by hydrocarbon dissociation processes, the poloidal profile of the ion flux is significantly changed. However the change in the ion flux at the target is small. The weak rollover of the ion flux appears due to the combination of effects of the radial diffusion loss and the wall pumping.

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m 2 . Dependence of the heat flux mitigation on a D 2 gas-puff is evaluated by SONIC simulations for high density ( n e_ave ∼ 1 × 10 20 m −3 ) high current plasmas. It is found that the peak heat load 10 MW/m 2 with dense ( n ed > 4 × 10 20 m −3 ) and cold ( T ed , T id ⩽ 1 eV) divertor plasmas are obtained at a moderate gas-puff of Γ puff = 15 × 10 21 s −1 . Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m 3 /s. In full non-inductive current drive plasmas with low density ( n e_ave ∼ 5 × 10 19 m −3 ), the reduction of divertor heat load is achieved with the Ar injection.

The PARASOL code and the simulation by using PARASOL are introduced briefly. The PARASOL code with particle-in-cell (PIC) method and binary collision model was developed in JAERI and JAEA. Simulations using PARASOL code were carried out in order to investigate the power and particle control with diveror system in fusion reactors. The one-dimensional (1D) version of PARASOL was adopted to investigate the Bohm criterion, the supersonic flow, the SOL heat conduction, and so on. The heat propagation due to edge localized mode (ELM) was studied with the 1D-dynamic PARASOL. The two-dimensional version of PARASOL for the whole tokamak plasma including scrape-off-layer (SOL)-divertor region was useful for simulating the SOL flow pattern, the electric field formation etc. Based on PARASOL simulation results, improved physics modeling for the fluid simulation was built up.

The accumulation process of high-Z impurity in toroidal rotating tokamak plasma is investigated. A new inward pinch of high-Z impurity due to the ionization/recombination processes is derived using an analytic model. This inward pinch is driven by the large deviation of a drift orbit from a magnetic surface and the resultant variation of the charge state along the drift orbit. The pinch velocity increases with increasing toroidal rotation speed in both the co-direction and the ctr-direction. The inward pinch expected by the analytic model is really observed in the numerical simulation using the IMPGYRO code (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Parallel electron heat flux in the SOL is determined via fully kinetic simulations using the one-dimensional particle-in-cell code PARASOL. The heat flux in fluid simulations is usually approximated in the collisional limit by the Spitzer-Harm heat flux qSH and in the collisionless limit by the one-way free-streaming heat flux qFS adjusted by the coefficient αe. Though αe is defined at the collisionless limit, in practice it is used generally and taken to be a constant ∼0.1. We survey the dependence of αe with respect to many plasma parameters: collisionality, source and sink relative position, source model, and sink radiation rate. Increasing radiation and the size of the radiation region generally increases αe. Without Langevin heating and electron radiation, αe is found tobe non-monotonic with collisionality. This effect is traced to the fact that in a collisionless plasma, the electron velocity distribution exhibits a high-energy tail, which implies that use of the one-way free-streaming heat flux is inaccurate. A new treatment for the free-streaming heat flux is proposed, wherein the symmetric bulk electron distribution is ignored, which makes αe ∼ 1 at all radiation rates for moderate and low collisionality. This method does not work for Langevin heating, which maintains a Maxwellian electron energy distribution (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The plasma flow in the scrape-off layer (SOL) plays an important role for the control of heat and particle including impurity in magnetic fusion reactors. SOL flow patterns have recently been studied by the particle simulations, and the effects of finite-orbit-size of ions are found to be essential for the flow-pattern formation [Takizuka et al., Nucl. Fusion 49, 075038 (2009)]. Based on these simulation results, a new model of the edge plasma flow is developed by introducing the “ion-orbit-induced flow” to the fluid equations. A tokamak plasma is divided into three regions; core region, transition layer and SOL region. The “ion-orbit-induced flow” is modeled by separating untrapped part and trapped part, and by taking account the collision effect and poloidal distribution. The “ion-orbit-induced flow” becomes large at the edge region (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Release conditions of a spherical dust particle from a plasma-facing vertical wall immersed in an oblique magnetic field are studied analytically. The Poisson′s equation is solved in the magnetic pre-sheath and electrostatic Debye sheath to obtain the electric field at the wall, which repels the dust particle from the wall. The electric field decreases as the direction of the magnetic field approaches parallel to the wall. On the other hand the ion flow velocity becomes larger at moderate oblique angle than that without magnetic field. It is clarified that in the case of the strong magnetic field or the low plasma density the critical dust radius for the release of the dust increases as magnetic field becomes more acute. In the case of the weak magnetic field or the high plasma density, the critical radius disappears at the moderate angles of the oblique magnetic field. The surface roughness widely changes the angle between the magnetic field lines and the surface normal.

Using an impurity Monte-Carlo code, IMPMC, the kinetic effect of thermal force on the He transport is investigated for JT-60SA detached plasmas. Without the recycling process, the kinetic effect of the thermal force is found to increase the He density in the divertor region by a factor of ∼2, compared with the conventional (fluid) evaluation. However, the kinetic effect is masked by the recycling at the target plates.

The modification of JT-60 to a fully superconducting coil tokamak, JT-60SA (JT-60 Super Advanced) device, has been programmed to contribute and supplement ITER toward to DEMO. Lower divertor design with the ITER-like lower single null divertor configuration is studied to obtain high heat and particle controllability using SOLDOR/NEUT2D code. With anticipated total power flux into SOL of 37 MW (90% of input power), the peak heat load on outer divertor target can be reduced to 5.8 MW/m 2 at the detached condition by gas puffing in the vertical divertor target with the “V-shaped corner”. It is ∼2/5 of the allowable level of 15 MW/m 2 . On the other hand, particle controllability such as control of detached to attached condition by divertor pumping is improved by increase the strike point distance from 20 to 120 mm with above divertor geometry, suggesting that recover from severe detachment at the small distance case can be achieving by elevation of the strike point locations. Optimization of upper divertor design is in progress for high β steady-state operation using upper single null divertor configuration.

The effect of the pressure profile on the energy loss caused by edge localized modes (ELMs) has been investigated by using an integrated simulation code TOPICS-IB based on a core transport code with a stability code for the peeling-ballooning modes and a transport model for scrape-off-layer and divertor plasmas. The steep pressure gradient inside the pedestal top is found to broaden the region of the ELM enhanced transport through the broadening of eigenfunctions and enhance the ELM energy loss. The ELM energy loss in the simulation becomes larger than 15 % of the pedestal energy, as is shown in the database of multi-machine experiments. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

In order to understand the particle pumping on JT-60U, we analyze the roles of atomic and molecular processes using SOLDOR/NEUT2D code. A case of short strike point distance shows that most of neutrals produced on the targets go toward the exhaust slot directly. Whereas, neutrals are scattered in the spherically at random for the long distance case by collision processes and a few of them go toward the slot. It is clarified that the incident neutrals to the slot at low ne/high Te divertor plasma condition are dominated by atoms. Those at high ne/low Te condition are dominated by molecules due to elastic collision. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Effects of oblique magnetic field on release condition of a spherical dust particle from a plasma-facing vertical wall is studied analytically. The magnetic pre-sheath and the Debye sheath are coupled to obtain the plasma quantities and the electric field at the wall, which are necessary to analyze the release condition. It is clarified that for the deeper potential drop inside the Debye sheath than the floating one the critical radius for release increases as the magnetic field approaches parallel to the wall. The smaller dust than the critical one can be released from the wall. On the other hand in the case of the shallower potential drop the critical radius disappears at some angle of the oblique magnetic field. From this analysis we find that the size of the released dust particle can be controllable by adjusting the plasma parameters such as plasma density and temperature as well as biasing the wall potential. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Taken impurity radiation and momentum loss in the divertor region into consideration, the applicability of Core-SOL-Divertor(CSD) model to the divertor detachment is investigated. Three solutions for the low recycling state, the high recycling state and the detached state can be obtained from the CSD model. The dependence of the ratio of the decay length for the density (Δn) and temperature (ΔT) in the SOL region is also examined, and it is found that Δn/ΔT ∼ T0.5s scaled by the SOL temperature (Ts), which is based on the experimental data, has a good potential to reproduce B2-EIRENE results. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A simulation study of the divertor for JT-60SA is discussed in both double-null (DN) and single-null (SN) configurations. In the DN case, the transition of peak heat load on the targets and divertor plasma states from CDN (connected double-null) to DDN (disconnected double-null) are shown in the simulation. The radiative cooling power in the divertor plasma is discussed in this study. The carbon impurity generated by the sputtering on the divertor target is included in the simulation. In the gas puffing cases of fueling gas (D2) and impurity gas (Ne), the reduction of heat load is confirmed consistently with increasing the radiative cooling loss power in the diverter plasma. The loss power and the distribution of radiative cooling in the divertor plasma are studied in this paper. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The dynamics of SOL-divertor plasmas after an ELM crash is studied with a one-dimensional particle simulation code, PARASOL. The ELM crash occurs off-centrally in the SOL region, and the ELM heat flux to the near divertor plate and that to the far divertor plate become asymmetric. The peak heat flux to the near plate is larger as compared to the far plate. The asymmetry in the peak heat flux increases with the connection-length ratio. The imbalance in the heat deposition, however, is small. The radial transport loss of ELM flux creates the asymmetry in the heat deposition, but the imbalance is still not large even for the large radial transport loss rate. The electron heat flux to the far plate brought by the SOL current is one of the causes of a small imbalance in the heat deposition. Another cause is the asymmetric SOL flow and its convective heat flux, whose stagnation point stays for a long period near the ELM crash location. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Electron cyclotron current drive (ECCD) has been numerically simulated in magnetic islands caused by neo-classical tearing modes. The electron drift orbits are tracked with the Coulomb collisions and the quasi-linear diffusion by EC waves, which are simulated by the Monte-Carlo method. The EC resonance region is assumed to be located around the O-point and localized in the toroidal direction. The driven current is well confined in the helical flux tube which includes the EC resonance region. The driven current channel looks like a 'snake' in real space. As the results of the ECCD with 10 MW in a plasma with ne = 3 × 1019 m−3 and Te = 10 keV, the current drive efficiency is about 1.6 times higher than that of an axi-symmetric plasma with no magnetic island. The driven current profile tends to peak around the O-point with increasing EC wave power.

Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6–8.3, βN values of 2.1–2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5–7, βN values of 2.3–3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% non-inductive current.

An integrated simulation code TOPICS-IB based on a transport code with a stability code for the peeling–ballooning modes and a scrape-off-layer (SOL) model has been developed to clarify self-consistent effects of edge localized modes (ELMs) and the SOL on the plasma performance. Experimentally observed collisionality dependence of the ELM energy loss is found to be caused by both the edge bootstrap current and the SOL transport. The bootstrap current decreases with an increase in collisionality and intensifies the magnetic shear at the pedestal region. The increase in the magnetic shear reduces the width of eigenfunctions of unstable modes, which results in the reduction of both the area of the ELM enhanced transport and the ELM enhanced transport near the separatrix. On the other hand, when an ELM crash occurs, the energy flows into the SOL and the SOL temperature rapidly increases. The increase in the SOL temperature lowers the ELM energy loss due to the flattening of the radial edge gradient. The parallel electron heat conduction determines how the SOL temperature increases. For higher collisionality, the conduction becomes lower and the SOL electron temperature increases more. By the above two mechanisms, the ELM energy loss decreases with increasing collisionality.

To predict the heat and particle controllability in the divertor of tokamak reactors and to optimize the divertor design, comprehensive simulations by integrated modeling allowing for various physical processes are indispensable. SOL/divertor codes have been developed in the Japan Atomic Energy Agency for the interpretation and the prediction of behaviour of SOL/divertor plasmas, neutrals and impurities. The code system consists of the two-dimensional fluid code SOLDOR, the neutral Monte-Carlo (MC) code NEUT2D and the impurity MC code IMPMC. Their integration code 'SONIC' is almost completed and examined to simulate self-consistently the SOL/divertor plasmas in JT-60U. In order to establish the physics modelling used in fluid simulations, the particle simulation code PARASOL has also been developed.Simulation studies using those codes have progressed with the analysis of JT-60U experiments and the divertor designing of JT-60SA (modification program of JT-60U). The X-point multifaceted asymmetric radiation from the edge in the JT-60U experiment is simulated. It is found that the deep penetration of chemically sputtered carbon at the dome causes the large radiation peaking near the X-point. The pumping capability of JT-60SA is evaluated through the simulation. A guideline to enhance the pumping efficiency is obtained in terms of the exhaust slot width and the strike point distance. Transient behaviour of SOL/divertor plasmas after an ELM crash is characterized by the PARASOL simulation; the fast-time-scale heat transport is affected by collisions while the slow-time-scale behaviour is affected by the recycling.

To characterize the divertor pumping for particle and heat control in the SOL/divertor and to extrapolate to future devices, simulations using the SOLDOR/NEUT2D code are performed on the JT-60U experiment. The simulation reproduces fairly well the neutral pressure and pumping flux in the exhaust chamber. A parametric survey shows that the pumping efficiency ηpump (ratio of pumping flux to generated flux around the divertor targets) increases with the pumping speed. It is found that a pumping speed higher than the present capability (26 m3/s) is necessary for the active particle control under the wall saturation condition. When the strike-point distance is shortened from 10 to 2 cm, the ηpump is enlarged by a factor of 1.5 with increase of the slot view angle. A virtual tilt of the divertor targets to 15° vertically enhances ηpump by a factor of 1.2 and leads simultaneously to a lower target heat load.

A dynamic five-point model is developed to study responses of scrape-off-layer (SOL) and divertor plasmas to a crash due to edge-localized modes (ELMs). The five-point model is beneficial to wide-ranging studies of SOL-divertor physics and to coupling with the core transport code in the integrated modeling. The five-point model can reproduce static and dynamic features obtained by usual fluid codes. The dynamic behavior obtained by the five-point model agrees fairly well with that by the particle code PARASOL. The influence of the ELM crash on the thermoelectric instability and the resultant asymmetry is investigated. The ELM crash transiently induces the thermoelectric instability and large SOL currents for asymmetric divertor plasmas before the ELM. The resultant currents drive convective heat flows and enhance the asymmetry on the fast time scale of electrons. The ELM crash is found to reverse the asymmetry before and after the ELM. Even for symmetric plasmas before the ELM, the asymmetry is induced on the slow time scale of ions by an ELM crash, because the transient formation of dense and cold divertor plasmas causes the thermoelectric instability.

In modeling of release of a dust particle from a plasma-facing wall it is usually assumed that electron velocity distribution is Maxwellian. However, the absorption of fast electrons by the conducting wall can lead to truncation of fast component of reflecting electrons from the wall. In this work we study the effect of truncation of electron velocity distribution on the release condition of a conducting spherical dust particle from the plasma-facing wall. The truncation increases the electric field at the wall surface compared to that calculated in absence of the truncation. The stronger electric field makes the dust particle hard released when the gravitational force is directed from the wall and applied wall potential is shallower than the floating one.