Table of contents

The National Ignition Facility (NIF) will allow scientists to prove the feasibility of inertial confinement fusion (ICF). The success of ICF experiments at NIF will critically depend on the availability of robust targets. Guided by computer simulations, we generated a new target design that takes advantage of the extreme atomic density of synthetic diamond, and developed a process that allows us to produce large quantities of these ultrahigh precision diamond targets via a low-cost batch process. Computer simulations were used to assess the performance and the robustness of these diamond targets. The results demonstrate that diamond has the potential to outperform other target materials in terms of energy efficiency and implosion stability, thus making successful ignition more likely.

Tokamak plasmas become less tolerant to externally applied non-axisymmetric magnetic 'error' fields as beta increases, due to a resonant interaction of the non-axisymmetric field with a stable n = 1 kink mode. Similar to observations in low beta plasmas, the limit to tolerable n = 1 magnetic field errors in neutral beam injection heated H-mode plasmas is seen as a bifurcation in the torque balance, which is followed by error field-driven locked modes and severe confinement degradation or a disruption. The error field tolerance is, therefore, largely determined by the braking torque resulting from the non-axisymmetric magnetic field. DIII-D experiments distinguish between a resonant-like torque, which decreases with increasing rotation, and a non-resonant-like torque, which increases with increasing rotation. While only resonant braking leads to a rotation collapse, modelling shows that non-resonant components can lower the tolerance to resonant components. The strong reduction of the error field tolerance with increasing beta, which has already been observed in early high beta experiments in DIII-D (La Haye et al1992 Nucl. Fusion32 2119), is linked to an increasing resonant field amplification resulting from a stable kink mode (Boozer 2001 Phys. Rev. Lett.86 5059). The amplification of externally applied n = 1 fields is measured with magnetic pick-up coils at normalized beta values as low as 1 and seen to increase with beta. The rate at which the amplification increases with beta becomes larger above the no-wall ideal MHD stability limit, where kinetic effects stabilize the resistive wall mode. The extent of the beta dependence and its importance for low torque scenarios was not previously appreciated, and was not included in the empirical scaling of the error field tolerance for ITER, which focused on the lowest density phase of a discharge prior to H-mode access (Buttery et al1999 Nucl. Fusion39 1827, 1999 ITER Physics Basis Nucl. Fusion39 2137). However, the measurable increase in the plasma response with beta can be exploited for 'dynamic' correction (i.e. with slow magnetic feedback) of the amplified error field.

The flow velocities of deuterons and low charge-state carbon ions have been measured simultaneously in the main scrape-off-layer (SOL) in low-density plasmas in DIII-D, and the dependences of these flow fields on the direction of the cross-field drifts (E × B and B × ∇B) have been investigated. These measurements were taken poloidally localized in the SOL region vertically opposite the divertor X-point. The carbon ion flows do not necessarily match those of the deuterons either in the direction with respect to the magnetic field lines or in magnitude, suggesting that physics effects apart from entrainment play a significant role in the impurity response. In configurations with the ion B × ∇B drift towards the divertor X-point, the parallel-B deuteron velocities at the plasma crown are high (−20 to −30 km s⁻¹ in the direction of the high field side (HFS) divertor), while they are nearly zero in configurations with the opposite B × ∇B drift direction. The flow direction of singly and doubly charged carbon ions is independent of the ion B × ∇B drift direction, and the ions flow at approximately −5 to −10 km s⁻¹ towards the HFS divertor. Simulations with the UEDGE code have been carried out to better understand the underlying physics processes. Inclusion of cross-field drifts in the simulations produced divertor solutions for density and temperature that agree significantly better with measured divertor parameters. These simulations do not, however, reproduce the measured flow fields at the crown for the configuration with the ion B × ∇B drift towards the divertor X-point. The UEDGE code has also been used to understand the influence of pumping at the HFS divertor plate, and a poloidal dependence in the radial transport coefficient.

Operating experimental devices have provided key inputs to the design process for ITER axisymmetric control. In particular, experiments have quantified controllability and robustness requirements in the presence of realistic noise and disturbance environments, which are difficult or impossible to characterize with modelling and simulation alone. This kind of information is particularly critical for ITER vertical control, which poses the highest demands on poloidal field system performance, since the consequences of loss of vertical control can be severe. This work describes results of multi-machine studies performed under a joint ITPA experiment (MDC-13) on fundamental vertical control performance and controllability limits. We present experimental results from Alcator C-Mod, DIII-D, NSTX, TCV and JET, along with analysis of these data to provide vertical control performance guidance to ITER. Useful metrics to quantify this control performance include the stability margin and maximum controllable vertical displacement. Theoretical analysis of the maximum controllable vertical displacement suggests effective approaches to improving performance in terms of this metric, with implications for ITER design modifications. Typical levels of noise in the vertical position measurement and several common disturbances which can challenge the vertical control loop are assessed and analysed.

In DIII-D, experiments have been performed in hydrogen plasmas to determine the requirement for hydrogen operation in ITER. The H-mode threshold power has been determined to increase with input torque for both hydrogen and deuterium plasmas with the H-mode power threshold for hydrogen plasmas being greater by approximately a factor of 2 at zero torque than in comparable deuterium plasmas. The threshold power for hydrogen discharges with full counter-current beam injection is roughly the same as the threshold power for deuterium discharges with co-current beam injection. The plasma geometry also influences the power threshold through the vertical distance between the X-point and the divertor surface.

In the Large Helical Device (LHD), direct oblique launching of the fundamental extraordinary (X-) mode from the high magnetic field side (HFS) is available without installation of any additional launching equipment on the inner side wall of the torus. In the experiment, power absorption was observed in two separated regions by the X-mode launching. The central electron density was about 8% of the cutoff density. The result of numerical analysis with the ray-tracing calculation suggests that most power of the launched X-mode is damped out as the X-mode in the fundamental electron cyclotron resonance (ECR) layer before it reaches the upper hybrid resonance (UHR) layer where the electron Bernstein wave (EBW) occurs which is excited via the slow X (SX-) B mode conversion process. Only about 0.2% of the launched power is mode converted to the EBW and is then absorbed at a maximum. One of the two separated power absorption regions observed in the experiment agrees well with the power absorption region of the X-mode suggested by the ray tracing. The other one agrees well with that of the O-mode despite the setting of the X-mode launching. Supposedly mixed waves of the X- and the O-mode might be launched in the experiment. We assumed that the incident transverse electromagnetic waves in vacuum couple with the electromagnetic modes in the plasma at the last closed flux surface (LCFS). However, the coupling point was supposedly located outside the LCFS.

For the few rays that can reach the UHR layer we have recognized that the parallel component of the refractive index N_∥ becomes close to zero and power absorption as the X-mode weakens when the rays pass through the fundamental ECR layer. A numerical investigation assuming a higher central electron density, that is 23% of the cutoff density, suggests a scenario of effective EBW excitation. The ray that passes through the centre of the focused Gaussian beam launched from the HFS so that N_∥ becomes close to zero near the ECR layer, Z can reach the UHR layer without being damped out and excite EBW. About 71% of power of the launched X-mode is mode converted to the EBW and absorbed in the Doppler-shifted ECR layer.

Observation of the parametric decay waves suggests that the decay wave was excited in the 'exterior' UHR layer which is located outside the LCFS before the launched X-mode reaches the HFS. The combination of tunnelling, reflection and mode-coupling processes, the so-called 'Budden's problem', is suggested to occur in the evanescent region between the 'exterior' right handed cyclotron cutoff and the UHR layer outside the LCFS. The tunnelling rate should be considered for estimating the power that can penetrate inside the LCFS from the HFS, in particular in the higher density regime where the excitation of the EBW is expected.

At temperatures below 400 °C, irradiation often causes hardening and reduction of elongation as well as toughness degradation to a considerable degree. Data, however, indicate that these changes remain in manageable ranges for ITER-TBM application. Moreover, the saturation tendency of these changes with neutron dose suggests that some of the reduced activation ferritic/martensitic steels are feasible even for future DEMO applications. It is also stressed that the development of a design methodology that is compatible with the large irradiation induced property changes is essential to enable these applications. Modelling activities for the macroscopic mechanical response are expected to play key roles in design methodology development. Macroscopic models of plasticity (a constitutive equation) and cyclic softening behaviour after irradiation are discussed. The significance of the models for estimating microstructural change during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.

Rotations in both poloidal and toroidal directions of a tokamak edge plasma have important interactions with various other plasma phenomena including plasma stability and transport. Solutions for rotation velocities were studied, using differential equations comprising the ambipolarity constraint and the parallel momentum balance equations of the revisited neoclassical theory, with the corrected contribution also from the gyro-viscosity tensor. Temperature and density profiles with realistic pedestal forms were considered given and controlled parametrically. The similarity of this equation system to reaction–diffusion equations was utilized in the numerical simulation and the study of critical points on the bifurcation diagrams. It was found that the steepness of the density and temperature gradients has important effects on the rotation stability and on its bifurcative behaviour.

In order to understand the physics of the ELM trigger and determine the ELM size, the fast ELM dynamics of type I and grassy ELMs have been studied in JT-60U, using new fast diagnostics with high spatial and temporal resolutions such as a lithium beam probe (Δt ∼ 0.5 ms) and a charge exchange recombination spectroscopy (Δt ∼ 2.5 ms), which can measure the electron density and the ion temperature, respectively. The evolution of the ion pressure profile in the pedestal region has been evaluated for the first time by detailed edge profile measurements. Then, the dynamics of the density, the ion temperature and the ion pressure in the ELM cycle has been investigated. The co-rotating plasmas are compared with the counter (ctr)-rotating plasmas for the understanding of the toroidal rotation effects. Type I ELMs observed in co-rotating plasmas exhibit a larger and wider ELM affected area (Δn_ped/n_ped ∼ 30%, radial extent >15 cm) than ctr-rotating plasmas (Δn_ped/n_ped ∼ 20%, radial extent ∼10 cm). Just before a type I ELM crash, the pedestal ion pressure and its maximum gradient in co-rotating plasmas are 20% and 12% higher than those in ctr-rotating plasmas, respectively. It is found that the radial extent of the ion pressure gradient at the pedestal region in co-rotating plasmas is 14% wider than that in ctr-rotating plasmas. The experimental results suggest that the ELM size is connected with the structure of the plasma pressure in the whole pedestal region. As for the dynamics of grassy ELMs, the collapse of density pedestal is smaller (<20%) and narrower (∼5 cm) than those of type I ELMs, as observed in the collapse of the electron temperature pedestal. Thus, it is confirmed that both conductive and convective losses due to grassy ELMs are small.

The ratio between the heat diffusion coefficients parallel and perpendicular to the magnetic field lines, χ_||/χ_⊥, influences the flattening of the temperature profile inside magnetic islands and the driving term of neoclassical tearing modes (Fitzpatrick 1995 Phys. Plasmas2 825). The value of this anisotropy is, however, not easily accessible experimentally. This paper presents a method to determine it from a systematic comparison of temperature measurements at magnetic islands with numerical heat diffusion simulations. The application of the method is demonstrated for a 2/1 magnetic island in the TEXTOR tokamak, where a heat diffusion anisotropy of 10⁸ is observed. This is lower by a factor of 40 than predicted by Spitzer and Härm (Spitzer and Härm 1953 Phys. Rev.89 997) and a strong indication that the heat flux limit determines the flattening of the electron temperature across magnetic islands.

Rseduction of heat loading appropriate for the plasma facing components such as the divertor is crucial for a fusion reactor. Power handling by large radiative power loss has been studied in long pulse ELMy H-mode discharges on JT-60U (τ_d = 30–35 s). Case 1 is argon (Ar) seeding into standard ELMy H-mode plasmas, where large radiation loss in the confined region of the main plasma caused a change in ELM characteristics from Type-I to Type-III. Case 2 is a combination of Ar and nitrogen (Ne) gas seeding into Type-I ELMy H-mode plasmas with an internal transport barrier (ITB). For case 1, large radiation loss both from the main plasma and from the divertor was produced, and operation of Type-III ELMs was preferable to a reduction in ELM energy loss fraction (W_ELM/W_dia) to 0.15%. Both transient and steady-state heat loadings were reduced. Relatively good energy confinement (H_H98y2 = 0.87 − 0.75) with large f_rad (P_rad/P_abs > 0.8) and divertor plasma detachment was sustained continuously for 13.5 s. For case 2, with reduced Ar seeding to the main plasma and increased divertor radiation with Ne seeding, the ELMy H-mode plasma with an ITB had better energy confinement (H_H98y2 = 0.95 − 0.8), which was sustained continuously for 12 s. The radiated power was increased primarily in the divertor ( ), which was produced both by seeded Ne ions and by carbon influx due to transient (ELM) and steady-state heat loadings in the attached divertor. Reduction in the heat loading was not enough, thus enhancement of the radiated power in the divertor will be necessary for the formation of the divertor detachment.

Based on the recent experimental results of the LHD and the magnet technology–cost basis developed for the ITER construction, the design windows of helical reactors are analysed. For searching design windows and investigating their economical potential, we have developed a mass–cost estimating model linked with the system design code (HeliCos). We found that the LHD-type helical reactor has the technically and economically attractive design windows, where the major radius is increased as large as for the sufficient blanket space, but the magnetic stored energy is decreased to a reasonable level because of lower magnetic field with the convenient physics basis of the H factor near 1.1 to the ISS04 scaling and beta value of 5%.

We study the effect of a background of microturbulence generated by unstable short scale length electron temperature gradient (ETG) modes on the excitation and nonlinear evolution of neoclassical tearing modes (NTMs) in a tokamak. A self-consistent model calculation is carried out in which low frequency stresses generated by the ETG modes act on the NTM and which in turn acts back on the ETG evolution by modulating the turbulence profile. The principal physical consequences of the ETG turbulence are to create an anomalous current diffusivity and an anomalous resistivity contribution. These turbulent transport contributions are incorporated in the evolution equations of a single helicity NTM to obtain a generalized Rutherford equation, and their impact on the instability threshold and growth rate is discussed.

Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 °C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

The transition of ASDEX Upgrade (AUG) from a graphite device to a full tungsten device is demonstrated with a reduction by an order of magnitude in both the carbon deposition and deuterium retention. The tungsten source is dominated by sputtering from intrinsic light impurities, and the tungsten influxes from the outboard limiters are the main source for the plasma. In H-mode discharges, central heating (neutral beams, ECRH) is used to increase turbulent outward transport avoiding tungsten accumulation. ICRH can only be used after boronization as its application otherwise results in large W influxes due to light impurities accelerated by electrical fields at the ICRH antennas. ELMs are important in reducing the inward transport of tungsten in the H-mode edge barrier and are controlled by gas puffing. Even without boronization, stationary, ITER baseline H-modes (confinement enhancement factor from ITER 98(y, 2) scaling H₉₈ ∼ 1, normalized beta β_N ∼ 2), with W concentrations below 3 × 10⁻⁵ were routinely achieved up to 1.2 MA plasma current.

The compatibility of high performance improved H-modes with unboronized W wall was demonstrated, achieving H₉₈ = 1.1 and β_N up to 2.6 at modest triangularities δ ⩽ 0.3 as required for advanced scenarios in ITER. With boronization the light impurities and the radiated power fraction especially in the divertor were reduced and the divertor plasma was actively cooled by N₂ seeding. N₂ seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges. The energy confinement increased to H₉₈-factors of 1.25 (β_N ∼ 2.7) and thereby exceeded the best values in a carbon-dominated AUG machine under similar conditions. Recent investigations show that this improvement is due to higher temperatures rather than to peaking of the electron density profile.

Further ITER discharge scenario tests include the demonstration of ECRF assisted low voltage plasma start-up and current rise to q₉₅ = 3 at toroidal electric fields below 0.3 V m⁻¹, to achieve a ITER compatible range of plasma internal inductance of 0.71–0.97. The results reported here strongly support tungsten as a first wall material solution.

On the Alcator C-Mod tokamak, lower hybrid current drive (LHCD) is being used to modify the current profile with the aim of obtaining advanced tokamak (AT) performance in plasmas with parameters similar to those that would be required on ITER. To date, power levels in excess of 1 MW at a frequency of 4.6 GHz have been coupled into a variety of plasmas. Experiments have established that LHCD on C-Mod behaves globally as predicted by theory. Bulk current drive efficiencies, n₂₀I_lhR/P_lh ∼ 0.25, inferred from magnetics and MSE are in line with theory. Quantitative comparisons between local measurements, MSE, ECE and hard x-ray bremsstrahlung, and theory/simulation using the GENRAY, TORIC-LH CQL3D and TSC-LSC codes have been performed. These comparisons have demonstrated the off-axis localization of the current drive, its magnitude and location dependence on the launched n_∥ spectrum, and the use of LHCD during the current ramp to save volt-seconds and delay the peaking of the current profile. Broadening of the x-ray emission profile during ICRF heating indicates that the current drive location can be controlled by the electron temperature, as expected. In addition, an alteration in the plasma toroidal rotation profile during LHCD has been observed with a significant rotation in the counter-current direction. Notably, the rotation is accompanied by peaking of the density and temperature profiles on a current diffusion time scale inside of the half radius where the LH absorption is taking place.

The injected power required to induce a transition from L-mode to H-mode plasmas is found to depend strongly on the injected neutral beam torque and consequent plasma toroidal rotation. Edge turbulence and flows, measured near the outboard midplane of the plasma (0.85 < r/a < 1.0) on DIII-D with the high-sensitivity 2D beam emission spectroscopy (BES) system, likewise vary with rotation and suggest a causative connection. The L–H power threshold in plasmas with the ion ∇B drift directed away from the X-point decreases from 4–6 MW with co-current beam injection, to 2–3 MW near zero net injected torque and to <2 MW with counter-injection in the discharges examined. Plasmas with the ion ∇B drift directed towards the X-point exhibit a qualitatively similar though less pronounced power threshold dependence on rotation. 2D edge turbulence measurements with BES show an increasing poloidal flow shear as the L–H transition is approached in all conditions. As toroidal rotation is varied from co-current to balanced in L-mode plasmas, the edge turbulence changes from a uni-modal character to a bi-modal structure, with the appearance of a low-frequency (f = 10–50 kHz) mode propagating in the electron diamagnetic direction, similar to what is observed as the ion ∇B drift is directed towards the X-point in co-rotating plasmas. At low rotation, the poloidal turbulence flow near the edge reverses prior to the L–H transition, generating a significant poloidal flow shear that exceeds the measured turbulence decorrelation rate. This increased poloidal turbulence velocity shear appears to facilitate the L–H transition. No such reversal is observed in high rotation plasmas. The high-frequency poloidal turbulence velocity spectrum exhibits a transition from a geodesic acoustic mode zonal flow to a higher-power, lower frequency zero-mean-frequency zonal flow as rotation varies from co-current to balanced during a torque scan at constant injected neutral beam power, perhaps also facilitating the L–H transition. This reduced power threshold at lower toroidal rotation may benefit inherently low-rotation plasmas such as ITER.

As it was recognized that local electron cyclotron (EC) wave power losses can be a competitive contribution to the 1D electron power balance for reactor-grade tokamak plasmas in regimes as anticipated for steady-state operation, a systematic effort is ongoing to improve the modelling capability for the radial profile of EC wave emission. This effort aims at generating a hierarchy of codes that cover the non-local behaviour of EC wave transport for inhomogeneous plasmas and in the presence of reflecting walls with increasingly improved accuracy and also provide sufficient computational efficiency for being usable in 1D transport studies. The recently developed code RAYTEC, which explicitly addresses the geometrical effects present in toroidal plasmas with arbitrary cross-section, is described and used to investigate the impact of elongation of the plasma cross-section and of toroidicity on the angular dependence of the EC radiation field, on the profile of the net EC wave power density lost from the plasma and on the total EC power loss for ITER-like plasma conditions. Furthermore, a comparison is made with the results of simpler models in use to describe both local and total EC power losses as well as with ones obtained from analytical formulae that are introduced on the basis of Trubnikov's formula for EC power emission.

We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a free-boundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; β_N above 3, H₉₈(y, 2) up to 2.0, f_BS up to 0.76 and f_NI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of q_min is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work. Finally, ideal MHD stability is investigated for the ITER-relevant advanced scenarios in KSTAR. The methods and results presented in this paper are expected to contribute to improving the ITER and beyond ITER predictive simulations.

This paper describes recent upgrades of TOPICA (Torino Politecnico Ion Cyclotron Antennas) formulation and implementation. TOPICA is a code capable of handling both the actual geometry of ion cyclotron (IC) antennas and an accurate plasma description; it can predict the performances of IC launchers with an unprecedented accuracy and numerical efficiency, validated against data measured in plasma operation conditions. In the reported upgrade, a new multi-cavity approach is introduced, producing significant savings in terms of CPU and memory requirements and allowing the analysis of large antennas even with limited computational resources. In fact, by formally separating the structure's cavities with a number of mathematical surfaces called 'apertures', the method of moments interaction matrix is block-wise sparse and, as a consequence, can be manipulated with a far higher numerical efficiency; this also allows an out-of-core solution.

The new approach of integrating magnetic field line trajectories in natural canonical coordinates (Punjabi and Ali 2008 Phys. Plasmas15 122502) in divertor tokamaks is used for the DIII-D tokamak (Luxon and Davis1985 Fusion Technol.8 441). The equilibrium EFIT data (Evans et al 2004 Phys. Rev. Lett.92 235003, Lao et al 2005 Fusion Sci. Technol.48 968) for the DIII-D tokamak shot 115467 at 3000 ms is used to construct the equilibrium generating function (EGF) for the DIII-D in natural canonical coordinates. The EGF gives quite an accurate representation of the closed and open equilibrium magnetic surfaces near the separatrix, the separatrix, the position of the X-point and the poloidal magnetic flux inside the ideal separatrix in the DIII-D. The equilibrium safety factor q from the EGF is somewhat smaller than the DIII-D EFIT q profile. The equilibrium safety factor is calculated from EGF as described in the previous paper (Punjabi and Ali 2008 Phys. Plasmas15 122502). Here the safety factor for the open surfaces in the DIII-D is calculated. A canonical transformation is used to construct a symplectic mapping for magnetic field line trajectories in the DIII-D in natural canonical coordinates. The map is explored in more detail in this work, and is used to calculate field line trajectories in the DIII-D tokamak. The continuous analogue of the map does not distort the DIII-D magnetic surfaces in different toroidal planes between successive iterations of the map. The map parameter k can represent effects of magnetic asymmetries in the DIII-D. These effects in the DIII-D are illustrated. The DIII-D map is then used to calculate stochastic broadening of the ideal separatrix from the topological noise and field errors, the low mn, the high mn and peeling–ballooning magnetic perturbations in the DIII-D. The width of the stochastic layer scales as 1/2 power of amplitude with a maximum deviation of 6% from the Boozer–Rechester scaling (Boozer and Rechester 1978 Phys. Fluids21 682). The loss of poloidal flux scales linearly with the amplitude of perturbation with a maximum deviation of 10% from linearity. Perturbations with higher mode numbers result in higher stochasticity. The higher the complexity and coupling in the equilibrium magnetic geometry, the closer is the scaling to the Boozer–Rechester scaling of width. The comparison of the EGF for the simple map (Punjabi et al 1992 Phys. Rev. Lett.69 3322) with that of the DIII-D shows that the more complex the magnetic geometry and the more coupling of modes in equilibrium, the more robust or resilient is the system against the chaos-inducing, symmetry-breaking perturbations.

Global electrostatic ITG turbulence physics, together with background dynamics, has been simulated in a realistic tokamak core geometry using XGC1, a full-function 5D gyrokinetic particle code. An adiabatic electron model has been used. Some verification exercises of XGC1 have been presented. The simulation volume extends from the magnetic axis to the pedestal top inside the magnetic separatrix. Central heating is applied, and a number, momentum and energy conserving linearized Monte Carlo Coulomb collision is used. In the turbulent region, the ion temperature gradient profile self-organizes globally around R/L_T = (Rd logT/dr = major radius on the magnetic axis/temperature gradient length) ≃6.5–7, which is somewhat above the conventional nonlinear criticality of ≃6. The self-organized ion temperature gradient profile is approximately stiff against variation of heat source magnitude. Results indicate that the relaxation to a self-organized state proceeds in two phases, namely, a transient phase of excessively bursty transport followed by a 1/f avalanching phase. The bursty types of behaviour are allowed by the quasi-periodic collapse of local E × B shearing barriers.

The large range of plasma currents (I_p = 0.2–1.6 MA) and feedback-controlled magnetic boundary conditions of the RFX-mod experiment make it well suited to performing scaling studies. The assessment of such scaling, in particular those on temperature and energy confinement, is crucial both for improving the operating reversed-field pinch (RFP) devices and for validating the RFP configuration as a candidate for the future fusion reactors. For such a purpose scaling laws for magnetic fluctuations, temperature and energy confinement have been evaluated in stationary operation. RFX-mod scaling laws have been compared with those obtained from other RFP devices and numerical simulations. The role of the magnetic boundary has been analysed, comparing discharges performed with different active control schemes of the edge radial magnetic field.

Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

It has been observed that with increasing toroidal rotation velocity inside the q = 1 layer in the direction opposite to the plasma current, sawtooth activity becomes moderate and tungsten accumulation becomes significant. The tungsten accumulation level is significantly reduced from this trend in the case when electron cyclotron wave or high energy neutral beam is injected into the plasma core. In contrast, the tungsten accumulation is kept high by the electron cyclotron wave injection into the peripheral region.

X-ray bursts from nonthermal electrons (∼10 keV to 1 MeV) emanating from the core plasma and the limiter are measured simultaneously in runaway dominated discharges in the SINP tokamak using a tangentially viewing NaI(Tl) detector and a detector that views the limiter areas. The radial transport coefficient of the nonthermal electrons has been derived from the cross-correlation function of the two x-ray intensities. Two phases of the plasma discharges with distinct transport of the nonthermal electrons are found in these experiments. The stochastic magnetic field fluctuations obtained from the transport coefficients are found to have a relationship with the autocorrelation functions of the x-ray emissivity signals.

A new method for determining the temporal evolution of plasma rotation is reported in this work. The method is based upon the detection of two different portions of the spectral profile of a plasma impurity line, using a monochromator with two photomultipliers installed at the exit slits. The plasma rotation velocity is determined by the ratio of the two detected signals. The measured toroidal rotation velocities of C III (4647.4 Å) and C VI (5290.6 Å), at different radial positions in TCABR discharges, show good agreement, within experimental uncertainty, with previous results (Severo et al2003 Nucl. Fusion43 1047). In particular, they confirm that the plasma core rotates in the direction opposite to the plasma current, while near the plasma edge (r/a > 0.9) the rotation is in the same direction. This technique was also used to investigate the dependence of toroidal rotation on the poloidal position of gas puffing. The results show that there is no dependence for the plasma core, while for plasma edge (r/a > 0.9) some dependence is observed.

DIII-D experiments have investigated ITER startup scenarios, including an initial phase where the plasma was limited on low field side poloidal bumper limiters. In addition, l_i feedback control has been tested with the goal of producing discharges in ITER within the capabilities of the poloidal field coil set and favourable to the intended mode of operations in the subsequent constant current (flattop) phase. These discharges have been modelled using the Corsica free boundary equilibrium code. High performance hybrid scenario discharges (β_N = 2.8, H_98,y2 = 1.4) and ITER H-mode baseline discharges (β_N > 1.6, H_98,y2 = 1–1.2) have been obtained experimentally in an ITER similar shape after the ITER-relevant startup. Studies have been initiated to develop a reliable scenario for exiting the burn phase and ramping down the plasma current in ITER without disruptions.

Properties of the toroidal momentum diffusivity (χ_ϕ) and the convection velocity (V_conv) in JT-60U H-mode plasmas have been obtained by plasma parameter scans such as the plasma current, neutral beam heating power and electron density. The toroidal momentum diffusivity increases with increasing heat diffusivity (χ_i) over a wide range of radii (r/a = 0.2–0.6) and χ_ϕ/χ_i ∼ 0.7–3 at the half radius (r/a = 0.5). The inward convection velocity (−V_conv) increases with increasing χ_ϕ, and −V_conv/χ_ϕ ∼ 0.5–2 (m⁻¹) at r/a = 0.5. It is found that the ratio χ_ϕ/χ_i increases with increasing ion temperature and decreases with increasing electron density. These tendencies are observed in other radial positions of r/a = 0.3, 0.4 and 0.6. Moreover, the ratio −V_conv/χ_ϕ at r/a = 0.4, 0.5 and 0.6 increases with increasing ion and electron temperatures or temperature gradients.