Showing papers in "Nuclear Fusion in 2011"

A first-principles predictive model of the pedestal height and width: development, testing and ITER optimization with the EPED model

Abstract: We develop and test a model, EPED1.6, for the H-mode pedestal height and width based upon two fundamental and calculable constraints: (1) onset of non-local peeling–ballooning modes at low to intermediate mode number, (2) onset of nearly local kinetic ballooning modes at high mode number. Calculation of these two constraints allows a unique, predictive determination of both pedestal height and width. The present version of the model is first principles, in that no parameters are fit to observations, and includes important non-ideal effects. Extensive successful comparisons with existing experiments on multiple tokamaks, including experiments where predictions were made prior to the experiment, are presented, and predictions for ITER are discussed.

Neutron-induced transmutation effects in W and W-alloys in a fusion environment

Abstract: W and W-alloys are among the primary candidate materials for plasma-facing components in the design of fusion reactors, particularly in high-heat-flux regions such as the divertor. Under neutron irradiation W undergoes transmutation to its near-neighbours in the periodic table. Additionally He and H are particles emitted from certain neutron-induced reactions, and this is particularly significant in fusion research since the presence of helium in a material can cause both swelling and a strong increase in brittleness. This paper presents the results of inventory burn-up calculations on pure W and gives quantitative estimates for He production rates in both a fusion-reactor environment and under conditions expected in the ITER experimental device. Transmutation reactions in possible alloying elements (Re, Ta, Ti and V), which could be used to reduce the brittleness of pure W, are also considered. Additionally, for comparison, the transmutation of other fusion-relevant materials, including Fe and SiC, are presented.

Survey of disruption causes at JET

Abstract: A survey has been carried out into the causes of all 2309 disruptions over the last decade of JET operations. The aim of this survey was to obtain a complete picture of all possible disruption causes, in order to devise better strategies to prevent or mitigate their impact. The analysis allows the effort to avoid or prevent JET disruptions to be more efficient and effective. As expected, a highly complex pattern of chain of events that led to disruptions emerged. It was found that the majority of disruptions had a technical root cause, for example due to control errors, or operator mistakes. These bring a random, non-physics, factor into the occurrence of disruptions and the disruption rate or disruptivity of a scenario may depend more on technical performance than on physics stability issues. The main root cause of JET disruptions was nevertheless due to neo-classical tearing modes that locked, closely followed in second place by disruptions due to human error. The development of more robust operational scenarios has reduced the JET disruption rate over the last decade from about 15% to below 4%. A fraction of all disruptions was caused by very fast, precursorless unpredictable events. The occurrence of these disruptions may set a lower limit of 0.4% to the disruption rate of JET. If one considers on top of that human error and all unforeseen failures of heating or control systems this lower limit may rise to 1.0% or 1.6%, respectively.

Disruption mitigation by massive gas injection in JET

Abstract: Disruption mitigation is mandatory for ITER in order to reduce forces, to mitigate heat loads during the thermal quench and to avoid runaway electrons (REs). A fast disruption mitigation valve has been installed at JET to study mitigation by massive gas injection. Different gas species and amounts have been investigated with respect to timescales and mitigation efficiency. We discuss the mitigation of halo currents as well as sideways forces during vertical displacement events, the mitigation of heat loads by increased energy dissipation through radiation, the heat loads which could arise by asymmetric radiation and the suppression of REs.

Performance of different tungsten grades under transient thermal loads

Abstract: Plasma facing components in future thermonuclear fusion devices will be subjected to intense transient thermal loads due to type I edge localized modes (ELMs), plasma disruptions, etc. To exclude irreversible damage to the divertor targets, local energy deposition must remain below the damage threshold for the selected wall materials. For monolithic tungsten (pure tungsten and tungsten alloys) power densities above ?0.3?GW?m?2 with 1?ms duration result in the formation of a dense crack network. Thin tungsten coatings for the so-called ITER-like wall in JET, which have been deposited on a two-directional carbon?fibre composite (CFC) material, are even less resistant to thermal shock damage; here the threshold values are by a factor of 2 lower. First ELM-simulation experiments with high cycle numbers up to 104 cycles on actively cooled bulk tungsten targets do not reveal any cracks for absorbed power densities up to 0.2?GW?m?2 and ELM-durations in the sub-millisecond range (0.8?ms); at somewhat higher power densities (0.27?GW?m?2, ?t = 0.5?ms) cracks have been detected for 106 cycles.

The reconstruction and research progress of the TEXT-U tokamak in China

Abstract: The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 1019 m−3, and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

Benchmarking of the mono-energetic transport coefficients - Results from the International Collaboration on Neoclassical Transport in Stellarators (ICNTS)

Abstract: Numerical results for the three mono-energetic transport coefficients required for a complete neoclassical description of stellarator plasmas have been benchmarked within an international collaboration. These transport coefficients are flux-surface-averaged moments of solutions to the linearized drift kinetic equation which have been determined using field-line-integration techniques, Monte Carlo simulations, a variational method employing Fourier–Legendre test functions and a finite-difference scheme. The benchmarking has been successfully carried out for past, present and future devices which represent different optimization strategies within the extensive configuration space available to stellarators. A qualitative comparison of the results with theoretical expectations for simple model fields is provided. The behaviour of the results for the mono-energetic radial and parallel transport coefficients can be largely understood from such theoretical considerations but the mono-energetic bootstrap current coefficient exhibits characteristics which have not been predicted.

A fusion-driven subcritical system concept based on viable technologies

Abstract: A fusion-driven hybrid subcritical system (FDS) concept has been designed and proposed as spent fuel burner based on viable technologies. The plasma fusion driver can be designed based on relatively easily achieved plasma parameters extrapolated from the successful operation of existing fusion experimental devices such as the EAST tokamak in China and other tokamaks in the world, and the subcritical fission blanket can be designed based on the well-developed technologies of fission power plants. The simulation calculations and performance analyses of plasma physics, neutronics, thermal-hydraulics, thermomechanics and safety have shown that the proposed concept can meet the requirements of tritium self-sufficiency and sufficient energy gain as well as effective burning of nuclear waste from fission power plants and efficient breeding of nuclear fuel to feed fission power plants.

Real-time physics-model-based simulation of the current density profile in tokamak plasmas

Abstract: A new paradigm is presented to reconstruct the plasma current density profile in a tokamak in real-time. The traditional method of basing the reconstruction on real-time diagnostics combined with a real-time Grad–Shafranov solver suffers from the difficulty of obtaining reliable internal current profile measurements with sufficient spatial and temporal accuracy to have a complete picture of the profile evolution at all times. A new methodology is proposed in which the plasma current density profile is simulated in real-time by solving the first-principle physics-based equations determining its evolution. Effectively, an interpretative transport simulation similar to those run today in post-plasma shot analysis is performed in real-time. This provides real-time reconstructions of the current density profile with spatial and temporal resolution constrained only by the capabilities of the computational platform used and not by the available diagnostics or the choice of basis functions. The diagnostic measurements available in real-time are used to constrain and improve the accuracy of the simulated profiles. Estimates of other plasma quantities, related to the current density profile, become available in real-time as well. The implementation of the proposed paradigm in the TCV tokamak is discussed, and its successful use in plasma experiments is demonstrated. This framework opens up the possibility of unifying q profile reconstructions across different tokamaks using a common physics model and will support a wealth of applications in which improved real-time knowledge of the plasma state is used for feedback control, disruption avoidance, scenario monitoring and external disturbance estimation.

Overview of toroidal momentum transport

Abstract: Toroidal momentum transport mechanisms are reviewed and put in a broader perspective. The generation of a finite momentum flux is closely related to the breaking of symmetry (parity) along the field. The symmetry argument allows for the systematic identification of possible transport mechanisms. Those that appear to lowest order in the normalized Larmor radius (the diagonal part, Coriolis pinch, E x B shearing, particle flux, and up-down asymmetric equilibria) are reasonably well understood. At higher order, expected to be of importance in the plasma edge, the theory is still under development.

ITER predictions using the GYRO verified and experimentally validated trapped gyro-Landau fluid transport model

Abstract: The trapped gyro-Landau fluid (TGLF) transport model computes the quasilinear particle and energy driftwave fluxes in tokamaks with shaped geometry, finite aspect ratio and collisions. The TGLF particle and energy fluxes have been successfully verified against a large database of collisionless nonlinear gyrokinetic simulations using the GYRO code. Using a new collision model in TGLF, we find remarkable agreement between the TGLF quasilinear fluxes and 64 new GYRO nonlinear simulations with electron–ion collisions. In validating TGLF against DIII-D and JET H-mode and hybrid discharges we find the temperature and density profiles are in excellent agreement with the measured profiles. ITER projections using TGLF show that the fusion gains are somewhat more pessimistic than the previous GLF23 results primarily due to finite aspect ratio effects included only in TGLF. The synergistic effects of density peaking, finite β and E × B shear due to finite toroidal rotation lead to significant increases in fusion power above a reduced physics ITER base case. The TGLF results for ITER are confirmed using nonlinear GYRO simulations in place of TGLF to predict the temperature profiles within the TGYRO transport code. These results represent a snapshot of the ongoing effort to improve the TGLF model, validate it against experimental data, and make predictions for ITER.

Advances towards QH-mode viability for ELM-stable operation in ITER

Abstract: The application of static, non-axisymmetric, nonresonant magnetic fields (NRMFs) to high beta DIII-D plasmas has allowed sustained operation with a quiescent H-mode (QH-mode) edge and both toroidal rotation and neutral beam injected torque near zero. Previous studies have shown that QH-mode operation can be accessed only if sufficient radial shear in the plasma flow is produced near the plasma edge. In past experiments, this flow shear was produced using neutral beam injection (NBI) to provide toroidal torque. In recent experiments, this torque was nearly completely replaced by the torque from applied NRMFs. The application of the NRMFs does not degrade the global energy confinement of the plasma. Conversely, the experiments show that the energy confinement quality increases with lower plasma rotation. Furthermore, the NRMF torque increases plasma resilience to locked modes at low rotation. These results open a path towards QH-mode utilization as an edge-localized mode (ELM)-stable H-mode in the self-heated burning plasma scenario, where toroidal momentum input from NBI may be small or absent.

Modelling of plasma response to resonant magnetic perturbation fields in MAST and ITER

Abstract: The resonant magnetic perturbation (RMP) fields, including the plasma response, are computed within a linear, full toroidal, single-fluid resistive magnetohydrodynamic (MHD) model, and under realistic plasma conditions for MAST and ITER. The response field is found to be considerably reduced, compared with the vacuum field produced by the magnetic perturbation coils. This field reduction relies strongly on the screening effect from the toroidal plasma rotation. Computations also quantify three-dimensional (3D) distortions of the plasma surface, caused by RMP fields. A correlation is found between the computed mode structures, the plasma surface displacement and the observed density pump-out effect in MAST experiments. Generally, the density pump-out tends to occur when the surface displacement peaks near the X-points.

Study on H-mode access at low density with lower hybrid current drive and lithium-wall coatings on the EAST superconducting tokamak

Abstract: The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of HIPB98(y,2) ∼ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge. The threshold power for H-mode access follows the international tokamak scaling even in the low density range and a threshold in density has been identified. With increasing accumulation of deposited Li the H-mode duration was gradually extended up to 3.6 s corresponding to ∼30 confinement times, limited only by currently attainable durations of the plasma current flat top. Finally, it was observed that neutral density near the lower X-point was progressively reduced by a factor of 4 with increasing Li accumulation, which is considered the main mechanism for the H-mode power threshold reduction by the Li wall coatings. (Some figures in this article are in colour only in the electronic version)

ELM simulation experiments on Pilot-PSI using simultaneous high flux plasma and transient heat/particle source

Abstract: A new experimental setup has been developed for edge localized mode (ELM) simulation experiments with relevant steady-state plasma conditions and transient heat/particle source. The setup is based on the Pilot-PSI linear plasma device and allows the superimposition of a transient heat/particle pulse to the steady-state heat flux plasma. Energy densities as high as 1?MJ?m?2 have been reached for a pulse duration of about 1.5?ms, and for a variety of gases (H, He, Ar). In this contribution, we report on the first experiments investigating the effect of the combined steady-state/pulsed plasma on polycrystalline tungsten targets. Under such conditions the threshold for tungsten release and surface roughening is found to be much lower than in previously reported experiments. This suggests that the combination of the high flux plasma and transient heat/particle source leads to strong synergistic effects.

Consequences of profile shearing on toroidal momentum transport

Abstract: Turbulent transport of toroidal momentum is investigated in global linear gyrokinetic simulations. The poloidal tilt of the global mode structure arising from the radial variation of the equilibrium (profile shearing) is shown to induce non-diagonal non-pinch momentum transport (residual stress). Local simulations performed at finite radial wave vector show that the effect is mainly due to the antisymmetric radial component of the magnetic drift. The residual stress resulting from profile shearing enhances co-current rotation for ion temperature gradient turbulence and counter-current rotation for trapped electron mode turbulence. (Some figures in this article are in colour only in the electronic version)

Effect of He on D retention in W exposed to low-energy, high-fluence (D, He, Ar) mixture plasmas

Abstract: W targets are exposed at fixed temperature in the range ~420–1100 K, to either pure D2, D2–δHe (0.1 < δ < 0.25), or D2–δHe–γAr (γ = 0.03) mixture plasma, or He pretreatment plasma followed by exposure to D2 plasma. A strong reduction in D retention is found for exposure temperature above 450 K and incident He-ion fluence exceeding ~1024 m−2. Reduced D retention values lie well below that measured on D2 plasma-exposed reference targets, and the scatter in retention values reported in the literature. A small level of Ar admixture to D2–0.1He plasma, leading to an Ar ion density fraction of ~3%, is found to have minimal effect on the D inventory reduction caused by He. In targets with reduced inventory, nuclear-reaction analysis reveals shallow D trapping (<50 nm), in the same locale as nanometre-sized bubbles observed using transmission electron microscopy. It is suggested that near-surface bubbles grow and interconnect, forming pathways leading back to the plasma–material interaction surface, thereby interrupting transport to the bulk and reducing D retention.

Physics and engineering design of the accelerator and electron dump for SPIDER

Abstract: The ITER Neutral Beam Test Facility (PRIMA) is planned to be built at Consorzio RFX (Padova, Italy). PRIMA includes two experimental devices: a full size ion source with low voltage extraction called SPIDER and a full size neutral beam injector at full beam power called MITICA. SPIDER is the first experimental device to be built and operated, aiming at testing the extraction of a negative ion beam (made of H− and in a later stage D− ions) from an ITER size ion source. The main requirements of this experiment are a H−/D− extracted current density larger than 355/285 A m−2, an energy of 100 keV and a pulse duration of up to 3600 s.Several analytical and numerical codes have been used for the design optimization process, some of which are commercial codes, while some others were developed ad hoc. The codes are used to simulate the electrical fields (SLACCAD, BYPO, OPERA), the magnetic fields (OPERA, ANSYS, COMSOL, PERMAG), the beam aiming (OPERA, IRES), the pressure inside the accelerator (CONDUCT, STRIP), the stripping reactions and transmitted/dumped power (EAMCC), the operating temperature, stress and deformations (ALIGN, ANSYS) and the heat loads on the electron dump (ED) (EDAC, BACKSCAT).An integrated approach, taking into consideration at the same time physics and engineering aspects, has been adopted all along the design process. Particular care has been taken in investigating the many interactions between physics and engineering aspects of the experiment. According to the 'robust design' philosophy, a comprehensive set of sensitivity analyses was performed, in order to investigate the influence of the design choices on the most relevant operating parameters.The design of the SPIDER accelerator, here described, has been developed in order to satisfy with reasonable margin all the requirements given by ITER, from the physics and engineering points of view. In particular, a new approach to the compensation of unwanted beam deflections inside the accelerator and a new concept for the ED have been introduced.

Lithium?metal infused trenches (LiMIT) for heat removal in fusion devices

Abstract: Observation of liquid lithium flow in metal trenches has been made using a lithium–metal infused trench (LiMIT) tile and is reported here The flow is self-pumping and uses thermoelectric magnetohydrodynamics to remove heated lithium and replenish it at a lower temperature Flow velocities have been measured and compared with theoretical predictions

L-mode validation studies of gyrokinetic turbulence simulations via multiscale and multifield turbulence measurements on the DIII-D tokamak

Abstract: A series of carefully designed experiments on DIII-D have taken advantage of a broad set of turbulence and profile diagnostics to rigorously test gyrokinetic turbulence simulations. In this paper the goals, tools and experiments performed in these validation studies are reviewed and specific examples presented. It is found that predictions of transport and fluctuation levels in the mid-core region (0.4 < ρ < 0.75) are in better agreement with experiment than those in the outer region (ρ ≥ 0.75) where edge coupling effects may become increasingly important and multiscale simulations may also be necessary. Validation studies such as these are crucial in developing confidence in a first-principles based predictive capability for ITER.

Effects of 3D magnetic perturbations on toroidal plasmas

Abstract: Small three-dimensional (3D) magnetic field perturbations have many interesting and possibly useful effects on tokamak and quasi-symmetric stellarator plasmas. Plasma transport equations that include these effects, most notably on diamagnetic-level toroidal plasma flows, have recently been developed. The 3D field perturbations and their plasma effects can be classified according to their toroidal mode number n: low n (say 1–5) resonant (with field line pitch, q = m/n) and non-resonant fields, medium n (~20, due to toroidal field ripple) and high n (due to microturbulence). Low n non-resonant fields induce a neoclassical toroidal viscosity (NTV) that damps toroidal rotation throughout the plasma towards an offset rotation in the counter-current direction. Recent tokamak experiments have generally confirmed and exploited these predictions by applying external low n non-resonant magnetic perturbations. Medium n toroidal field ripple produces similar effects plus possible ripple-trapping NTV effects and ion direct losses in the edge. A low n (e.g. n = 1) resonant field is mostly shielded by the toroidally rotating plasma at and inside the resonant (rational) surface. If it is large enough it can stop plasma rotation at the rational surface, facilitate magnetic reconnection there and lead to a growing stationary magnetic island (locked mode), which often causes a plasma disruption. Externally applied 3D magnetic perturbations usually have many components. In the plasma their lowest n (e.g. n = 1) externally resonant components can be amplified by kink-type plasma responses, particularly at high β. Low n plasma instabilities (e.g. resistive wall modes, neoclassical tearing modes) cause additional 3D magnetic perturbations in tokamak plasmas. Tearing modes in their nonlinear (Rutherford) regime bifurcate the topology and form magnetic islands. Finally, multiple resonant magnetic perturbations (RMPs) can, if not shielded by plasma rotation effects, cause local magnetic stochasticity and increase plasma transport in the edge of H-mode plasmas. These various effects of 3D fields can be used to modify directly the plasma toroidal rotation (and possibly transport via multiple RMPs for controlling edge localized modes) and indirectly anomalous plasma transport. The present understanding and modelling of these various 3D magnetic field perturbation effects including for test blanket modules in ITER are summarized. Finally, implications of the present understanding and key open issues for developing a predictive capability of them for ITER are discussed.

Metal impurity transport control in JET H-mode plasmas with central ion cyclotron radiofrequency power injection

Abstract: The scan of ion cyclotron resonant heating (ICRH) power has been used to systematically study the pump out effect of central electron heating on impurities such as Ni and Mo in H-mode low collisionality discharges in JET. The transport parameters of Ni and Mo have been measured by introducing a transient perturbation on their densities via the laser blow off technique. Without ICRH Ni and Mo density profiles are typically peaked. The application of ICRH induces on Ni and Mo in the plasma centre (at normalized poloidal flux rho = 0.2) an outward drift approximately proportional to the amount of injected power. Above a threshold of ICRH power of about 3 MW in the specific case the radial flow of Ni and Mo changes from inwards to outwards and the impurity profiles, extrapolated to stationary conditions, become hollow. At mid-radius the impurity profiles become flat or only slightly hollow. In the plasma centre the variation of the convection-to-diffusivity ratio upsilon/D of Ni is particularly well correlated with the change in the ion temperature gradient in qualitative agreement with the neoclassical theory. However, the experimental radial velocity is larger than the neoclassical one by up to one order of magnitude. Gyrokinetic simulations of the radial impurity fluxes induced by electrostatic turbulence do not foresee a flow reversal in the analysed discharges.

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

Abstract: A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

L–H transition studies on DIII-D to determine H-mode access for operational scenarios in ITER

Abstract: A comprehensive set of L–H transition experiments has been performed on DIII-D to determine the requirements for access to H-mode plasmas in ITER's first (non-nuclear) operational phase with H and He plasmas and the second (activated) operational phase with D plasmas The H-mode power threshold, PTH, was evaluated for different operational configurations and auxiliary heating methods for the different main ion species Helium plasmas have significantly higher PTH than deuterium plasmas at low densities for all heating schemes, but similar PTH as deuterium plasmas at high densities except for H-neutral beam injection-heated discharges, which are still higher Changes in PTH are observed when helium concentration levels in deuterium plasmas exceed 40% There is a strong dependence of PTH on the magnetic geometry in the vicinity of the divertor The trend of decreasing PTH with decreasing X-point height is observed for all of the main ion species irrespective of the heating method, which appears to indicate that there is a common physics process behind this effect for all of the ion species Helium and deuterium plasmas exhibit a significant increase in PTH for strong resonant magnetic perturbations The application of a local magnetic ripple of 3% from test blanket module mock-up coils did not change PTH in deuterium plasmas

Nonlinear ELM simulations based on a nonideal peeling–ballooning model using the BOUT++ code

Abstract: A minimum set of equations based on the peeling–ballooning (P–B) model with nonideal physics effects (diamagnetic drift, E × B drift, resistivity and anomalous electron viscosity) is found to simulate pedestal collapse when using the BOUT++ simulation code, developed in part from the original fluid edge code BOUT. Linear simulations of P–B modes find good agreement in growth rate and mode structure with ELITE calculations. The influence of the E × B drift, diamagnetic drift, resistivity, anomalous electron viscosity, ion viscosity and parallel thermal diffusivity on P–B modes is being studied; we find that (1) the diamagnetic drift and E × B drift stabilize the P–B mode in a manner consistent with theoretical expectations; (2) resistivity destabilizes the P–B mode, leading to resistive P–B mode; (3) anomalous electron and parallel ion viscosities destabilize the P–B mode, leading to a viscous P–B mode; (4) perpendicular ion viscosity and parallel thermal diffusivity stabilize the P–B mode. With addition of the anomalous electron viscosity under the assumption that the anomalous kinematic electron viscosity is comparable to the anomalous electron perpendicular thermal diffusivity, or the Prandtl number is close to unity, it is found from nonlinear simulations using a realistic high Lundquist number that the pedestal collapse is limited to the edge region and the ELM size is about 5–10% of the pedestal stored energy. This is consistent with many observations of large ELMs. The estimated island size is consistent with the size of fast pedestal pressure collapse. In the stable α-zones of ideal P–B modes, nonlinear simulations of viscous ballooning modes or current-diffusive ballooning mode (CDBM) for ITER H-mode scenarios are presented.

Runaway electron confinement modelling for rapid shutdown scenarios in DIII-D, Alcator C-Mod and ITER

Abstract: MHD simulations of rapid shutdown scenarios by massive particle injection in DIII-D, Alcator C-Mod and ITER are performed in order to study runaway electron (RE) transport during mitigated disruptions. The simulations include a RE confinement model using drift-orbit calculations for test particles. A comparison of limited and diverted plasma shapes is studied in DIII-D simulations, and improved confinement in the limited shape is found due to both spatial localization and reduced toroidal spectrum in the nonlinear MHD activity. C-Mod simulations compare shutdown scenarios in which impurity (Ar) fuelling is concentrated in the edge versus the core, and the confinement of REs in the core is maintained until the onset of the m = 1/n = 1 mode, which is delayed in the case of edge deposition, relative to core deposition. But, the overall RE loss fraction is 100% regardless of Ar fuelling profile. A comparison of simulations across the three devices points to a trend of increased RE confinement with increasing device size, wherein all REs are lost in C-Mod, all are confined in ITER, and a partial loss is observed in DIII-D. This trend is related to a reduction in the fluctuating field amplitude near the plasma edge during the thermal-quench-induced MHD activity. The result bodes poorly for RE mitigation strategies in ITER that rely on MHD deconfinement of REs.

Confinement of ‘improved H-modes’ in the all-tungsten ASDEX Upgrade with nitrogen seeding

Abstract: In ASDEX Upgrade the compatibility of improved H-modes with an all-W wall has been demonstrated. Under boronized conditions light impurities and the radiated power fraction in the divertor were reduced, requiring N seeding to cool the divertor plasma. The impurity seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges by up to 25%. The energy confinement increases to H98-factors up to 1.3 and thereby exceeds the best values in the carbon-dominated AUG at the same density and collisionality. This improvement is due to higher edge temperatures rather than to peaking of the electron density profile. Higher temperatures are reached at the pedestal top leading, via profile stiffness, to an increase in the total plasma pressure. There is no change to in the plasma core. The dilution at the plasma edge by nitrogen seems to play an important role since it allows higher ion temperatures at the same edge ion pressure as in the unseeded case. The dilution of the core plasma remains moderate.

On velocity space interrogation regions of fast-ion collective Thomson scattering at ITER

Abstract: The collective Thomson scattering (CTS) diagnostic proposed for ITER is designed to measure projected 1D fast-ion velocity distribution functions at several spatial locations simultaneously. The frequency shift of scattered radiation and the scattering geometry place fast ions that caused the collective scattering in well-defined regions in velocity space, here dubbed interrogation regions. Since the CTS instrument measures entire spectra of scattered radiation, many different interrogation regions are probed simultaneously. We here give analytic expressions for weight functions describing the interrogation regions, and we show typical interrogation regions of the proposed ITER CTS system. The backscattering system with receivers on the low-field side is sensitive to fast ions with pitch |p| = |v∥/v| 0.6–0.8. Additionally, we use weight functions to reconstruct 2D fast-ion distribution functions, given two projected 1D velocity distribution functions from simulated simultaneous measurements with the back- and forward scattering systems.

Recent progress in RF heating and long-pulse experiments on EAST

Abstract: New capabilities including radio-frequency (RF) powers and diagnostics have been developed since the last IAEA meeting. Divertor performance has been systematically assessed for both single null and double null configurations with normal and reversed BT directions to address asymmetries correlated with classic drift and the issues of heat load on divertor plates. 1 MW lower hybrid wave power injection has produced fully non-inductive diverted plasma discharges. Such discharges have been readily extended to well over 1 min duration. With the assistance of lower hybrid current drive (LHCD), EAST can be operated at 1 MA with both RTEFIT/Iso-flux or pre-programmed control algorithms. A later operation mode could produce an improved confinement plasma by increasing the ohmic power through regulating the loop voltage. Both direct electron and ion heating by ion cyclotron resonant frequency power have been observed for the first time on EAST. Intrinsic toroidal rotation and momentum study in ohmic and LHCD plasmas reveals the importance of the balance between neoclassical viscosity and neutral friction at the edge and finds clear linkage between edge and core toroidal rotation. The expected capabilities on EAST in the next two years will allow access to high-performance regimes, such as H-modes, and enable specific physics studies.

An equilibrium model for tungsten fuzz in an eroding plasma environment

Abstract: A model equating the growth rate of tungsten fuzz on a plasma-exposed surface to the erosion rate of the fuzzy surface is developed to predict the likelihood of tungsten fuzz formation in the steady-state environment of toroidal confinement devices. To date this question has not been answered because the operational conditions in existing magnetic confinement machines do not necessarily replicate those expected in future fusion reactors (i.e. high-fluence operation, high temperature plasma-facing materials and edge plasma relatively free of condensable impurities). The model developed is validated by performing plasma exposure experiments at different incident ion energies (thereby varying the erosion rate) and measuring the resultant fuzz layer thickness. The results indicate that if the conditions exist for fuzz development in a steady-state plasma (surface temperature and energetic helium flux), then the erosion rate will determine the equilibrium thickness of the surface fuzz layer.