Core Plasma Confinement Research Articles

Observation of long-radial-range-correlation in turbulence in high-collisionality high-confinement plasmas on DIII-D

We report on the observation of spatially asymmetric turbulent structures with a long radial correlation length in the core of high-collisionality H-mode plasmas on DIII-D tokamak. These turbulent structures develop from shorter wavelength turbulence and have a radially elongated structure. The envelope of turbulence spans a broad radial range in the mid-radius region, leading to streamer-like transport events. The underlying turbulence is featured by intermittency, long-term memory effect, and the characteristic spectrum of self-organized criticality. The amplitude and the radial scale increase substantially when the shearing rate of the mean flow is reduced below the turbulent scattering rate. The enhanced long-radial-range-correlated (LRRC) transport events are accompanied by apparent degradation of normalized energy confinement time. The emergence of such LRRC transport events may serve as a candidate explanation for the degrading nature of H-mode core plasma confinement at high collisionality on DIII-D tokamak.

Characterization of mesoscopic turbulent transport events with long-radial-range correlation in DIII-D H-mode plasmas

A dimensionless collisionality scan has been performed in H-mode plasmas on DIII-D tokamak, with detailed measurements of intermediate-to-high wavenumber turbulence using Doppler backscattering systems. It is found that the shorter wavelength turbulence develops into spatially asymmetric turbulent structures with a long-radial-range correlation (LRRC) in the mid-radius region of high collisionality discharges. Linear cgyro simulations indicate that the underlying turbulence is likely driven by the electron-temperature-gradient mode. The LRRC transport events are highly intermittent and show a power spectrum of Sñ(k⊥)∝k⊥−1 for density fluctuations, which is often associated with self-organized criticality. The magnitude and the radial scale of those turbulent structures increase significantly when the Er×B mean flow shearing rate decreases at higher collisionality. The enhanced LRRC transport events appear to be correlated with the degraded energy confinement time. The emergence of such LRRC transport events may serve as a candidate explanation for the degrading nature of H-mode core plasma confinement at high collisionality.

The influence of full drifts on density shoulder formation at the midplane by numerical modeling

The density shoulder at the midplane may influence core plasma confinement during H-mode discharge, thus affecting long-pulse steady-state discharge. Drifts in the edge plasma play a remarkable role in plasma transport and the divertor operation regime, which determine density shoulder formation (DSF). In this work, the SOLPS-ITER code package is used to evaluate the influence of full drifts on DSF in poloidal and radial coordinates. An open divertor of DIII-D-like geometry with weak neutral compression is chosen for the modeling. Cases without drifts, with only E × B drifts in forward B t and with full drifts in both forward and reversed B t are simulated for comparison. It is confirmed that the high upstream density promotes DSF when the drift is not considered, which has also been observed in various investigations. When the drifts are taken into account, the divertor in/out asymmetry (or upstream ionization source) is determined by the direction of B t due to the variation of particle transport, thus the shoulder can be facilitated or suppressed. Two mechanisms of DSF with full drifts are elucidated: (1) E × B and B × ∇B drifts promote DSF at the inner midplane (IMP) by raising the ionization source (at IMP) in forward B t; (2) the drifts contribute to DSF at the outer midplane by enhancing the particle transport loss in reversed B t. In a high-recycling regime, ionization is the dominant term for DSF, while in the low-recycling regime enhanced particle transport loss plays a more important role. Comprehensively understanding the mechanisms of DSF is of great importance for the improvement of core–edge compatibility in fusion reactors.

Achievements of actively controlled divertor detachment compatible with sustained high confinement core in DIII-D and EAST

The compatibility of efficient divertor detachment with high-performance core plasma is vital to the development of magnetically controlled fusion energy. The joint research on the EAST and DIII-D tokamaks demonstrates successful integration of divertor detachment with excellent core plasma confinement quality, a milestone towards solving the critical plasma–wall-interaction (PWI) issue and core-edge integration for ITER and future reactors. In EAST, actively controlled partial detachment with T et,div ∼ 5 eV around the strike point and H 98 > 1 in different H-mode scenarios including the high β P H-mode scenario have been achieved with ITER-like tungsten divertor, by optimizing the detachment access condition and performing detailed experiments for core-edge integration. For active long-pulse detachment feedback control, a 30 s H-mode operation with detachment-control duration being 25 s has been successfully achieved in EAST. DIII-D has achieved actively controlled fully detached divertor with low plasma electron temperature (T et,div ⩽ 5 eV across the entire divertor target) and low particle flux (degree of detachment, DoD > 3), simultaneously with very high core performance (β N ∼3, β P > 2 and H 98 ∼ 1.5) in the high βP scenario being developed for ITER and future reactors. The high-β P high confinement scenario is characterized by an internal transport barrier (ITB) at large radius and a weak edge transport barrier (ETB, or pedestal), which are synergistically self-organized. Both the high-β P scenario and impurity seeding facilitate divertor detachment. The detachment access leads to the reduction of ETB, which facilitates the development of an even stronger ITB at large radius in the high β P scenario. Thus, this strong large radius ITB enables the core confinement improvement during detachment. These significant joint DIII-D and EAST advances on the compatibility of high confinement core and detached divertor show a great potential for achieving a high-performance core plasma suitable for long-pulse operation of fusion reactors with controllable steady-state PWIs.

Evaluation of core beta effects on pedestal MHD stability in ITER and consequences for energy confinement

The maximum stable pedestal pressure has been shown to increase with core pressure and, in combination with core transport effects, this can lead to a positive feedback mechanism. However, the effect is shown to saturate for a high β in ASDEX-Upgrade simulations [Wolfrum et al. “Impact of wall materials and seeding gases on the pedestal and on core plasma performance,” Nucl. Mater. Energy 12, 18 (2017)]. In this paper, it is numerically investigated whether this effect appears in ITER plasmas, using ideal MHD numerical codes HELENA and MISHKA for different ITER scenarios, in a range of plasma conditions: two inductive scenarios at 7.5 MA/2.65 T and 15 MA/5.3 T and one steady-state scenario at 10 MA/5.3 T. For all scenarios, reference cases for ITER plasmas were taken as a starting point. No pedestal pressure saturation is found for the inductive scenarios, gradually growing up to the global βN limit, which is determined by the Troyon limit. On the contrary, for the 10 MA/5.3 T steady-state scenario, the maximum stable pedestal pressure does not depend on the total β and it is limited by low-n kink-peeling modes, as opposed to high-n peeling-ballooning modes that limit the maximum attainable pedestal height in the inductive scenarios. This core-edge MHD stability feedback loop has been investigated for two assumptions regarding the scaling of the pedestal width with βp,ped¯, using either a constant pedestal width or when scaling it as ΔψN∝βp,ped¯1/2. A stronger core-edge MHD stability feedback is observed for the varying pedestal width for the inductive plasma scenarios, which is closer to the experimental results from JET [Challis et al. “Improved confinement in JET high plasmas with an ITER-like wall,” Nucl. Fusion 55(5), 053031 (2015)], but not for the steady-state one. Finally, the pressure achieved according to this core-edge feedback stability analysis is compared to the plasma pressure achievable on the basis of the energy confinement IPB98(y,2) scaling for various assumptions regarding the scaling of core plasma confinement with heating power.

On the Influence of the SOL Region on Core-Plasma Energy Confinement Time in Simulation of Turbulent Transport Processes in Tokamaks

The influence of the scrape-off layer (SOL) on the energy confinement time of the core plasma is analyzed by using computer simulation of the self-consistent evolution of turbulence and anomalous transport. The simulation is performed for the conditions of a series of discharges with auxiliary heating in the T‑10 tokamak. It is demonstrated that the role of the SOL region in core-plasma confinement increases considerably relative to the conventional diffusion model of transport processes in the presence of turbulent convection that maintains self-consistency of pressure profiles in plasma. In this case, the influence of the SOL region on the energy confinement time of the core plasma is modeled by nonlinear boundary conditions of the third type at the boundary between the core plasma and the SOL region in which the heat fluxes at the SOL boundary are expressed in the form of power-law dependences on local values of the electron and ion temperatures. The specific form of this power-law dependence is related to power-law scaling for effective plasma confinement time in the SOL. The results of numerical simulation of turbulent-plasma evolution revealed that such an approach allows to provide self-consistent evolution of the energy confinement time τE in transient regimes with auxiliary heating, resulting in τE attaining stationary values close to those observed in real experiments.

Core plasma confinement during detachment transition with RMP application in LHD

The core plasma confinement during detachment phase is investigated in the discharges with application of resonant magnetic perturbation (RMP) field in LHD. The RMP application creates a remnant magnetic island in the edge stochastic layer, which largely changes the plasma parameter profiles including impurity radiation. The electron temperature and pressure profiles are flattened at the island, while the electron density is slightly peaked at the edge of the island. The estimated impurity radiation profile is enhanced and fixed around the magnetic island during the detached phase, where the discharge is stably sustained with controlled level of radiation. Without RMP, the radiation penetrates the confinement region, leading to radiation collapse. It is found that in the case of the RMP application the plasma stored energy increases discontinuously at the detachment transition. In spite of the reduced effective plasma volume caused by the edge magnetic island and by the enhanced radiation there, the central plasma pressure finally exceeds the case without RMP. This is caused by the pressure profile peaking at the central region in the case with RMP. These results indicate clear change of core plasma confinement during the detached phase with RMP.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (ne/nGW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Control of 3D edge radiation structure with resonant magnetic perturbation fields applied to the stochastic layer and stabilization of radiative divertor plasma in LHD

It is found that resonant magnetic perturbation (RMP) fields have a stabilizing effect on the radiating edge plasma, realizing stable sustainment of radiative divertor (RD) operation in the Large Helical Device (LHD). Without RMP, thermal instability leads to radiative collapse. Divertor power load is reduced by a factor of 3–10 during the RMP-assisted RD phase, while maintaining relatively good core plasma confinement with confinement enhancement factor . It has also been demonstrated that the RMP field itself can initiate transition to RD operation by increasing perturbation strength, while keeping constant density and injection power. The results show a possibility of a new control knob for divertor power load in a 3D magnetic field configuration. It is also found that after the transition to RD, the energy confinement enhancement factor based on ISS04 scaling increases by a factor of 1.4 compared with the attached phase. The operation range of the RMP-assisted RD is identified in terms of RMP strength and radial location of the resonance layer of the RMP. A 3D edge radiation structure is analysed using the edge transport code EMC3-EIRENE and the results are compared with experiments. The comparison indicates that the application of RMP modulates the 3D edge radiation structure such that an intense radiation appears around the X-point of the m/n = 1/1 island in the case with RMP, while it is located at the inboard side without RMP.

Analyses of Core Plasma Confinement and Internal Transport Barrier Formation in Tokamak and Helical Reactors

Analyses of Core Plasma Confinement and Internal Transport Barrier Formation in Tokamak and Helical Reactors

Integrated modelling of a JET type-I ELMy H-mode pulse and predictions for ITER-like wall scenarios

A baseline type-I ELMy H-mode JET tokamak discharge in low triangularity has been analysed using the JINTRAC integrated code suite to obtain a self-consistent description of the plasma edge dynamics and core plasma confinement. The inter- and intra-ELM transport model of JINTRAC has been adapted as such to match the experimental pre- and post-ELM plasma pedestal profiles and at the same time to recover the observed ELM dynamics in terms of ELM frequency, ELM energy loss, target ELM wetted area and target heat flows. The presented and validated modelling results for the JET all-carbon device reference case are utilized to predict a type-I ELMy H-mode for the JET ITER-like wall (ILW) assuming a full-tungsten divertor and beryllium main-chamber wall using JINTRAC. By keeping all relevant transport parameters fixed for the inter- and intra-ELM phase as in the all-carbon reference case it is observed that a moderate amount of seeded neon impurity is necessary to compensate for a similar level of radiation when carbon is absent in the system. The results of the ILW model setup are finally taken to estimate the total amount of tungsten particles eroded per ELM from the target plates. A rough estimate of the core radiative fraction due to W accumulation is given, predicting that no substantial impact on typical a JET ILW type-I ELMy H-mode discharge core plasma performance is expected.

Bolometric imaging of detached plasmas in LHD

Detached plasmas can increase the radiated power fraction and thereby reduce the heat load on divertor tiles, which is important for steady-state fusion reactor operation. In LHD, divertor detachment is achieved through several means including gas puffing and the application of an m/n=1/1 magnetic island using external perturbation coils. A recently upgraded imaging bolometer with a tangential view provides unique images of the plasma radiation, which give essential information about the source, three-dimensional (3-D) structure and dynamics of the radiation belt in the edge region. When compared to the results of edge modelling (EMC3/EIRENE) they can verify the model assumptions and identify the radial location of the radiation region, which is important because it impacts the compatibility between heat load and core plasma confinement and is dependent on impurity transport and magnetic topological effects on energy transport which are included in the 3-D edge model.

Internal Transport Barrier Analysis Including Impurities in Tokamak and Helical Reactor Plasmas

The operation with Internal Transport Barrier (ITB) is expected as a high performance operation. ITB is utilized to improve core plasma confinement in the reversed magnetic shear. It is considered that the changes of core plasma profile by the ITB cause changes of impurity transport. In a large fusion reactor, high-Z materials will be used as plasma facing components because high loads of heat and particles concentrate there. However, high-Z impurities from these components cause large radiation loss and dilute the fuel even if the amount of impurities is small. Therefore, in this study, firstly, the ITB formation which includes the effects of the magnetic shear and perturbed profiles by the pellet injection was simulated using the Toroidal Transport Analysis Linkage code TOTAL. Secondly, we analyzed transport of the tungsten impurities using an impurity model in TOTAL code, and compared the impurity profile in the case with ITB to the one without ITB in the tokamak reactor. The impurities decreased in the ITB formation region when ITB was formed, and the outward flux of total impurity density was observed there. It can be expected that outward flux of impurities is generated by the temperature and the density gradients.

Core magnetic islands and plasma confinement in the H-1NF heliac

Plasma confinement in the vicinity of vacuum magnetic islands near the magnetic axis in the H-1NF heliac [S. M. Hamberger et al., Fusion Technol. 17, 123 (1990)] has been experimentally studied in a low temperature argon plasma. Experimental results indicate that, under favorable conditions, these low order (m=2) islands near the core of the plasma serve as “pockets” of higher electron density. This results in significant profile modifications including enhancement of the core radial electric field to a large positive value, possibly through an electron-root ambipolar condition. The characteristics of islands are found to be dependent on the plasma collisionality and island width.

Analysis of Internal Transport Barrier Formation in Tokamak Reactor Plasmas

Since a high performance operation is needed in the future fusion reactor, it is expected that the operation with Internal Transport Barrier (ITB) is utilized to improve core plasma confinement in the reversed magnetic shear. In this study, firstly, the simulation results are compared with the experimental result, and the validity of the model is checked. Secondly, the effects of the magnetic shear and the pellet injection on ITB formation are simulated in tokamak reactor using the Toroidal Transport Analysis Linkage code TOTAL. In this simulation, the ITB formation is confirmed in the condition that magnetic shear is nearly zero and the density gradient exists, and it found that the reversed magnetic shear configuration is effective in the operation with ITB. c

Measurement of edge currents in DIII-D and their implication for pedestal stability

The present performance limits of tokamak discharges are strongly coupled to the stability and transport properties of the edge plasma. Both experimental and modeling efforts have shown a clear connection between the edge pressure pedestal height and core plasma confinement. The key to understanding the stability and performance limits of the pedestal revolves around an accurate knowledge of the plasma current in this region. Using the Zeeman effect in an injected 30keV lithium beam, we have measured the currents in the edge of the DIII-D [J. L. Luxon, Nucl. Fusion 42, 6114 (2002)] tokamak for various confinement modes. This method of determining j(r) is insensitive to the large electric fields which coexist in the pedestal region and which complicate motional Stark effect measurements. For the high confinement cases, where substantial pedestal pressures exist, we find large (~MA∕m2), localized (ΔR∼1–2cm) currents in the pedestal region, located near the maximum in the pressure gradient. These values are consistent with calculations of edge bootstrap current using the neoclassical NCLASS [W. A. Houlberg, K. C. Shaing, S. P. Hirshman, and M. C. Zarnstorff, Phys. Plasmas 4, 3230 (1997)] and Sauter [O. Sauter, C. Angioni, and Y. R. Lin-Lin, Phys. Plasmas 6, 2834 (1999)] models and the measured pedestal density and temperature profiles. The apparent consistency of the measured jEDGE with neoclassical predictions occurs despite the violation of one of the fundamental tenets of the theory, namely, ε=ρi∕LP⪡1, where ρi is the ion poloidal gyroradius and Lp is the pressure gradient scale length. The measured jEDGE has also been used to generate self-consistent reconstructions using the free boundary equilibrium solvers CORSICA [T. A. Casper, T. B. Kaiser, R. A. Jong, L. L. LoDestro, J. Moller, and L. D. Pearlstein, Plasma Phys. Controlled Fusion 45, 1193 (2003)] and EFIT [L. L. Lao, H. E. St. John, R. D. Stambough, A. G. Kellman, and W. Pfeiffer, Nucl. Fusion 25, 1611 (1985)]. These equilibria allow us, in conjunction with the edge localized instabilities in tokamak experiments [P. B. Snyder, H. R. Wilson, J. R. Ferron, L. L. Lao, A. W. Leonard, T. H. Osborne, A. D. Turnbull, D. Mossessian, M. Murakami, and X. Q. Xu, Phys. Plasmas 9, 2037 (2002); H. R. Wilson, P. B. Snyder, G. T. A. Huysmans, and R. L. Miller, Phys. Plasmas 9, 1277 (2002)] magnetohydrodynamic stability code, to assess the linear stability of the edge to peeling∕ballooning modes. These results are then compared to the measured edge localized mode onset conditions and again good agreement is found between the experimental and model limits on the maximum permissible jEDGE.

Edge localized mode control with an edge resonant magnetic perturbation

A low amplitude (δbr∕BT=1 part in 5000) edge resonant magnetic field perturbation with toroidal mode number n=3 and poloidal mode numbers between 8 and 15 has been used to suppress most large type I edge localized modes (ELMs) without degrading core plasma confinement. ELMs have been suppressed for periods of up to 8.6 energy confinement times when the edge safety factor q95 is between 3.5 and 4. The large ELMs are replaced by packets of events (possibly type II ELMs) with small amplitude, narrow radial extent, and a higher level of magnetic field and density fluctuations, creating a duty cycle with long “active” intervals of high transport and short “quiet” intervals of low transport. The increased transport associated with these events is less impulsive and slows the recovery of the pedestal profiles to the values reached just before the large ELMs without the n=3 perturbation. Changing the toroidal phase of the perturbation by 60° with respect to the best ELM suppression case reduces the ELM amplitude and frequency by factors of 2–3 in the divertor, produces a more stochastic response in the H-mode pedestal profiles, and displays similar increases in small scale events, although significant numbers of large ELMs survive. In contrast to the best ELM suppression case where the type I ELMs are also suppressed on the outboard midplane, the midplane recycling increases until individual ELMs are no longer discernable. The ELM response depends on the toroidal phase of the applied perturbation because intrinsic error fields make the target plasma nonaxisymmetric, and suggests that at least some of the variation in ELM behavior in a single device or among different devices is due to differences in the intrinsic error fields in these devices. These results indicate that ELMs can be suppressed by small edge resonant magnetic field perturbations. Extrapolation to next-step burning plasma devices will require extending the regime of operation to lower collisionality and understanding the physical mechanism responsible for the ELM suppression.

Lightly- or non-seeded, partially detached ITER divertor operation

The ITER divertor scenario calls for the reduction of direct target loading to < 50 MW by the use of edge/divertor radiation, enhanced by a seeded recycling impurity, to ‘detach’ the divertor plasma from the target plates. At this radiated power fraction core plasma confinement quality is degraded, the radiation comes mainly from the X-point region and impurity control becomes difficult. In this paper we examine the possibility of using either no impurity seeding, or very light impurity seeding, in an attached mode of operation to alleviate these problems. It is concluded that such scenarios should be possible if an edge separatrix density of 5 × 10 19 m −3, which corresponds approximately to the ITER design value, can be achieved.

Characteristics of ergodic divertor plasmas in the Tore Supra tokamak

The characteristics of ergodic divertor (ED) deuterium plasmas in the Tore Supra tokamak are discussed. The divertor volume is a relatively well limited region between the ED modules and ρ = τ/a equivalent to 0.8. The core (ρ ⩽ 0.8) plasma confinement is not degraded by the presence of the ED (by comparison with similar limiter, non-ED plasmas): the electron density and temperature profiles are not modified. The ED configuration compares well with a classical X point divertor: intrinsic impurities are screened out, the power load is deflected to dedicated surfaces, and density control through particle pumping is possible. Up to 5.5 MW of ion cyclotron resonance power and 3 MW of lower hybrid power have been successfully used, power exhaust being mainly by radiation (80%)

Plasma-surface interactions in TFTR DT experiments

TFTR has begun its campaign to study deuterium-tritium fusion under reactor-like conditions. Variable amounts of deuterium and tritium neutral beam power have been used to maximize fusion power, study alpha heating, investigate alpha particle confinement, and search for alpha driven plasma instabilities. Additional areas of study include energy and particle transport and confinement, ICRF heating schemes for DT plasmas, tritium retention, and fusion in high β p plasmas. The majority of this work is done in the TFTR supershot confinement regime. To obtain supershots, extensive limiter conditioning using helium fueled ohmic discharges and lithium pellet injection into ohmic and neutral beam heated plasmas is performed, resulting in a low recycling limiter. The relationship between recycling and core plasma confinement has been studied by using helium, deuterium and high- Z gas puffs to simulate high recycling limiter conditions. These studies show that confinement in TFTR supershots is very sensitive to the influx of neutral particles at the plasma edge.