As a promising scenario for fusion reactors, the high poloidal-beta (beta(P)) scenario is characterized by a strong large radius internal transport barrier (ITB), which significantly enhances the overall confinement quality and the bootstrap current fraction for fully non-inductive operation. It is frequently observed that in the presence of a strong ITB, the pedestal height is lower and is accompanied by small edge localized modes (ELMs), which further improves the compatibility of a high performance core with an edge solution. A mechanism for the formation of the low pedestal is proposed in this paper. It is found that the strong ITB creates an off-axis bootstrap current to clamp the local safety factor q, and thus the magnetic shear in the outer core/pedestal region is increased. Gyrokinetic simulations with the CGYRO code show that the higher magnetic shear brings the experimental profiles into the range where the growth rate of drift-wave instabilities and thus transport is higher, and therefore a lower pedestal gradient is expected. The combination of low pedestal and high magnetic shear further enhances the turbulent transport across the whole pedestal, consistent with power balance analysis. Such a positive feedback mechanism ultimately results in a lower pressure pedestal as observed in experiments. Under such a low pedestal, linear simulations with BOUT++ predict the growth rates of peeling-ballooning modes to be lower across the whole toroidal mode number spectra, and the nonlinear BOUT++ simulation exhibits lower saturated fluctuation intensity as well, consistent with the experimentally observed lower ELM size.

Gyrokinetic simulation of a dedicated pedestal density ramping-up discharge on DIII-D can reproduce the enhancement of magnetic turbulence in the pedestal, which is identified to be caused by micro-tearing modes (MTMs). An increase of MTM amplitude results in higher electron thermal diffusivity, consistent with experimentally observed lower electron temperature gradient and degraded pedestal height. Gyrokinetic simulation identifies the major cause of MTM enhancement to be the increase of collisionality, which has a significant impact on the MTM intensity and is beyond the description of any (quasi-)linear theory.

Self-consistent modeling using the stability, transport, equilibrium, and pedestal (STEP) workflow in the OMFIT integrated modeling framework (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as China Fusion Engineering Test Reactor (CFETR) Zhuang et al (2019 Nucl. Fusion 59 112010) will benefit from high-density operation (Greenwald limit fraction fgw asymptotic to 0.7-1.3). Regimes with an operational density near the Greenwald limit will likely need peaked density profiles so that the pedestal density remains below the Greenwald limit. Peaked density profiles can be achieved with the help of pellet injection. A flexible Pellet Ablation Module (PAM), which predicts the density source based on a comprehensive analytical pellet ablation model, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. This workflow is applied to DIII-D and finds good agreement with experiments. On ITER the effect of pellet fueling is examined in an advanced inductive scenario, where a fusion gain of up to Q= 9 is predicted with strong central pellet fueling. On CFETR, with a mid-radius density source, an average of 1.5 x 1022 electrons s-1 are required to achieve the density and temperature profiles necessary for the 1000 MW advanced scenario with a tritium burn-up fraction of similar to 3%.

We report the observation of a set of coherent high frequency electromagnetic fluctuations that leads to a turbulence induced self-regulating phenomenon in the DIII-D high bootstrap current fraction plasma. The fluctuations have frequency of 130-220  kHz, the poloidal wavelength and phase velocity are 16-30  m^{-1} and ∼30  km/s, respectively, in the outboard midplane with the estimated toroidal mode number n∼5-9. The fluctuations are located in the internal transport barrier (ITB) region at large radius and are experimentally validated to be kinetic ballooning modes (KBM). Quasilinear estimation predicts the KBM to be able to drive experimental particle flux and non-negligible thermal flux, suggesting its significant role in regulating the ITB saturation.

Based on the theory of critical gradient model (CGM) and following the simulation method proposed by Waltz et al. [Nucl. Fusion 55, 123012 (2015)], a combination of TGLFEP and EPtran code is employed to predict the energetic particle (EP) transport induced by Alfvén eigenmodes (AEs). To be consistent with the experiment, recent improvements to the simulation method include consideration of threshold evolution and orbit loss due to finite orbit width. The revised CGM is applied to simulate two DIII-D experimental discharges (#142111 and #153071). It well reproduces the experimental profiles with multiple unstable AEs and large-scale EP transport. Discharge #142111 had previously been simulated using a nonlinear MHD-kinetic code MEGA [Todo et al., Nucl. Fusion 55, 073020 (2015)] with a transport mechanism based on stochasticity induced by overlapping AE. By comparing the simulated EP profiles, we find that the AE transport threshold is approximated by both the MEGA nonlinear stability threshold and the proposed CGM threshold (error <5% for single n and <17% for multiple n simulation). Both of them are larger than the linear stability threshold of the most unstable AE mode by a quantity of the order of the flux needed to sustain EP transport by the background turbulence. We have also applied the improved CGM to simulate the α particle redistribution for a China Fusion Engineering Test Reactor steady state scenario. Because of the clear separation between the AE unstable region and the loss cone, only a moderate α particle loss of ∼9.6% is predicted.

The edge transport and core accumulation of tungsten (W) particles on China Fusion Engineering Test Reactor (CFETR) have been studied by integrated modelling consisting of EMC3-EIRENE and STRAHL codes. The edge transport and power dissipation of W particles are simulated by EMC3-EIRENE. An in–out asymmetry of W (1–28)+ ions density has been revealed in the in- and out-board divertor regions. This is mainly due to the stronger reversal flow velocity of W ions at the outboard divertor. The upward flow of W ions near the separatrix leads to a moderate W impurity leakage from the divertor on CFETR compared to the existing full W device ASDEX Upgrade due to the high plasma density near the CFETR divertor targets. Further, the density distribution and radiation loss of W ions in the core region are investigated by STRAHL code. The high charge-state W (29–60)+ and W (61–74)+ ions mainly reside in the regions of Ψ N = 0.20–0.98 and 0.00–0.90 (Ψ N is the normalized poloidal magnetic flux), respectively. The W induced energy dissipation in different regions is assessed according to both STRAHL and EMC3-EIRENE simulations. Particularly, the impacts of the W core radiation on the operation regime are discussed according to the H-mode threshold scaling law proposed by Martin et al (2008 J. Phys.: Conf. Ser. 123 012033) for the baseline plasma on CFETR. Further, parameter studies on the pinch velocity ( v imp ) and diffusion coefficient ( D imp ) have been performed to check their impacts on the operation regime of CFETR. A three-fold increase of v imp / D imp results in a higher W core energy loss, which can lead to the transition from H-mode back to L-mode.

The DIII-D super-H (SH) scenario, which is characterized by a significantly higher pedestal pressure compared to standard high confinement mode (H mode) plasmas, typically exhibits two phases in its temporal dynamics. The early hot ion (SH–HI) phase has higher core ion temperatures and normalized confinement factor (H 98(y,2) ∼ 2) than the later ‘standard’ SH phase, which has similar pedestal pressure characteristics to the SH–HI phase but a lower confinement factor (H 98(y,2) ∼ 1.2) as well as lower pedestal T i/T e ratio. However, beyond the pedestal differences, it is also observed that in the core plasma T i is more peaked and has a significantly larger normalized gradient scale length a/L Ti in the SH–HI phase than in the SH phase. This paper identifies the physics responsible for the different core profiles via gyrokinetic and gyrofluid modeling. It is found that the ion temperature gradient (ITG) mode dominates the core transport for both phases. Absent flow shear effects, the ITG critical gradient (a/L Ti,crit) is shown to be far smaller in the SH–HI phase than the SH phase. The lower a/L Ti,crit in the SH–HI phase is shown to be mainly induced by the hollow carbon (impurity) density profile, which is strongly destabilizing relative to the nearly flat carbon density profile in the SH phase. Differences in the T i/T e ratio between these phases are found to have a minor impact. However, the significantly stronger flow shearing in the SH–HI phase relative to the SH phase enables the achievement of higher core a/L Ti values and is therefore mainly responsible for the higher core T i values observed in the early SH–HI phase. Predictive transport modeling shows that the confinement in the lower-rotation SH phase could be elevated significantly if a peaked impurity density profile can be achieved, and potential applications to the performance improvement of future reactors are discussed.

The numerical modelling of the heat flux distribution with neon impurity seeding on China fusion engineering test reactor has been performed by the three-dimensional (3D) edge transport code EMC3–EIRENE. The maximum heat flux on divertor targets is about 18 MW m −2 without impurity seeding under the input power of 200 MW entering into the scrape-off layer. In order to mitigate the heat loads below 10 MW m −2 , neon impurity seeded at different poloidal positions has been investigated to understand the properties of impurity concentration and heat load distributions for a single toroidal injection location. The majority of the studied neon injections gives rise to a toroidally asymmetric profile of heat load deposition on the in- or out-board divertor targets. The heat loads cannot be reduced below 10 MW m −2 along the whole torus for a single toroidal injection location. In order to achieve the heat load mitigation (<10 MW m −2 ) along the entire torus, modelling of sole and simultaneous multi-toroidal neon injections near the in- and out-board strike points has been stimulated, which indicates that the simultaneous multi-toroidal neon injections show a better heat flux mitigation on both in- and out-board divertor targets. The maximum heat flux can be reduced below 7 MW m −2 on divertor targets for the studied scenarios of the simultaneous multi-toroidal neon injections.

The EPED1 model and self-consistent core-pedestal coupling in integrated modeling are used to design the pedestal structure of the China Fusion Engineering Testing Reactor (CFETR) steady-state scenario. The key parameters, such as β p and q 95 , are based on the grassy edge-localized-mode (ELM) experimental database. In this work, we use the BOUT++ six-field two-fluid code to simulate the onset of the ELM in the CFETR steady-state scenario. The ELM size is around 0.2% in nonlinear simulations, which is in the experimental range of the grassy ELM discharges, 0.1%–1% observed in multiple tokamak devices. Linear and nonlinear simulations show that the dominant high- n ballooning modes peak around n = 40. Compared to type-I ELM crashing dynamics, grassy ELM crashing has a smaller initial crash and is then followed by three phases of turbulence spreading, which are dominated by multi-modes, a high- n mode of n = 45 and low- n mode of n = 5, respectively. In contras to type-I ELM, the perturbation of the high- n mode has a narrow width around ψ = 0.95, and magnetic island formation and reconnection occur only beyond ψ = 0.95, leading to a small initial crash. Mode–mode interaction in the multi-mode coexistence stage stops the growth of individual modes and reduces the transport of particles and heat, and these are the two reasons why the ELM size is small. In–out asymmetry of transient heat flux with a ratio of E out / E in = 3.5 is found during grassy ELM crash. The rise and delay times of the heat flux match the calculation from the free-streaming model. To evaluate the erosion of the divertor target, the energy fluence at the outer divertor target is calculated, which is 0.029 MJ m −2 , 5.5 times smaller than the tungsten melting limit 0.16 MJ m −2 . The calculated energy fluency still follows the experimental scaling law from type-I ELM experiments. The fluctuation eddies in the toroidal direction show a filament structure at the outer mid-plane. Parallel heat flux patterns with a toroidal mode number n = 10 are found at the outer divertor with an amplitude of 680 MW m −2 .

Babinov N. Bader A. Badziak J. Baek S. Bagryansky P. Bakharev N. Ball J. Bandyopadhyay I. Banerjee S. Banerjee S. Banon Navarro A. Bao J. Barbisan M. Barbui T. Bastiani S. Batani D. Batistoni P. Battaglia D. Baylor L. Becoulet M. Belokurov A. Belonohy E. Belova E. Bergmann A. Berk H. Beurskens M. Bhattacharjee A. Bhattacharyay R. Biel W. Bierwage A. Biewer T. Bilato R. Bin W. Bisai N. Bisson R. Bizarro J. Blanchard J. Boeglin W. Bolzonella T. Bombarda F. Bonanomi N. Bonnin X. Bonoli P. Boozer A. Borodin D. Borodkina I. Borthakur S. Bortolon A. Bosch H.-S. Bourdelle C. Breizman B. Bremer P.-T. Brezinsek S. Briguglio S. Brizard A. Brochard G. Bromberg L. Brookman M. Browning P. Brunetti D. Brunsell P. Bruzzone P. Bühler L. Bufferand H. Buller S. Buratti P. Burhenn R. Buzi L. Byggmästar J. Bykov I. Bykov V.

The critical gradient mode (CGM) is employed to predict the energetic particle (EP) transport induced by the Alfvén eigenmode (AE). To improve the model, the normalized critical density gradient is set as an inverse proportional function of energetic particle density; consequently, the threshold evolves during EP transport. Moreover, in order to consider the EP orbit loss mechanism in CGM, ORBIT code is employed to calculate the EP loss cone in phase space. With these improvements, the AE enhances EPs radial transport, pushing the particles into the loss cone. The combination of the two mechanisms raises the lost fraction to 6.6%, which is higher than the linear superposition of the two mechanisms. However, the loss is still far lower than that observed in current experiments. Avoiding significant overlap between the AE unstable region and the loss cone is a key factor in minimizing EP loss.

The study of edge localized mode (ELM) behavior, stationary heat flux and transient heat flux during the ELM crash phase is performed for a China Fusion Engineering Test Reactor (CFETR) 1 GW hybrid mode operation scenario ( R = 7.2 m, B T = 6.5 T, I p = 13.78 MA). Modeling and simulation start with a scenario obtained by multi-code integrated modeling on the one modeling framework for integrated tasks framework. Linear stability and nonlinear simulations of ELM dynamics are carried out using the BOUT++ six-field reduced magnetohydrodynamic module, which show a much smaller ELM energy loss (Δ ELM ~ 0.13%) compared to that of a Type-I ELM. Parametric analysis of the weak linear growth rate and small ELM energy loss characteristics shed light on physics corresponding to a grassy ELM regime for CFETR 1 GW hybrid scenario. The transient heat flux on the divertor target during this small ELM phase is investigated using BOUT++. We found that upstream radial transport in the scrape-off-layer (SOL) induced by small ELMs is weak, which keeps it in the drift-dominated region. However, the heat flux width is still broadened to λ q = 4.64 mm by the increase of separatrix temperature during ELM nonlinear evolution. The impact of transient peak heat load and ELM energy fluence on tungsten melting and net erosion rate of divertor target is evaluated for the first-time using physics-based transport that connects the pedestal with the SOL. Energy fluence caused by a single ELM pulse is below the tungsten melting limit, while tungsten erosion would exceed the material requirements. We conclude that external mitigation methods, such as divertor detachment and advanced divertor geometry are likely needed for the steady state operation of CFETR.

A conventional single null divertor geometry has been proposed for Chinese fusion engineering testing reactor (CFETR) with fusion power up to gigawatt level. Modeling using the SOLPS5.0 code package shows promising results for divertor power exhaust by seeding with neon (Ne) and deuterium (D-2). Partial detachment for both inner and outer divertor targets can be achieved with a peak heat load lower than 3MW/m(2). The effective ion charge number, Z(eff), at the core boundary is below three, which can be further reduced by increasing the upstream D-2 puffing rate. A higher D-2 puffing rate helps to increase electron density (n(e)) in scrape-off layer (SOL) and the impurity screening ability. Based on the SOLPS modeling results, the lifetime of tungsten (W) divertor targets has been estimated by using the DIVIMP code, which indicates that W sputtering is mainly contributed by Ne ions at a far SOL region due to high electron and ion temperature (T-e and T-i) there. Upstream D-2 puffing can reduce the W erosion rate and W impurity concentration inside the separatrix due to decreasing T-e and T-i at the far SOL region. The modeling results show a viable operation regime for the CFETR divertor.

We have identified a robust grassy-edge localized mode (ELM) operation regime for future tokamak reactors. The regime exists within a pedestal top electron collisionality (nu*) window at high global poloidal beta (beta(p)). The existence of an upper nu* limit for grassy-ELMs is consistent with results previously reported in experiments (Oyama et al 2010 Nucl. Fusion 50 064014), while the existence of a lower nu* limit has not been reported previously. Using EPED and BOUT + +, a theoretical model that quantitatively explains the physics of the grassy-ELMs within the window, which distinguishes them from the small mixed-ELMs at lower nu*, is presented for the first time. A peeling-ballooning stability boundary is obtained by scanning the operating density space. The change in density corresponds to a change in nu* that affects the pedestal bootstrap current. High beta(p) leads to a strong Shafranov shift, which affects the flux surface averaged pressure drive. The two effects combine to create a peeling-dominated window in intermediate nu* buffered by ballooning-dominated regimes. Only the peeling-dominated regime shows a cyclic behavior in the perturbed pressure during the nonlinear simulation of an ELM crash, reminiscent of grassy-ELM dynamics. Similarly, the energy released across the separatrix is demonstrated to be significantly smaller. The quick recovery of the ELM crash is explainable by the rapid rise of a low n kink-peeling instability when the pedestal current I-ped exceeds a threshold at high beta(p). It minimizes the excursion beyond marginal stability and is absent in the ballooning-dominated regime. Comparison with recent experiments over a range of beta(p) and nu* strongly supports the physical picture proposed by the modeling.

Investigation of turbulent transport dynamics in scrape-off-layer (SOL) and divertor heat flux width prediction is performed for ITER. Both BOUT ++ transport and BOUT ++ turbulence codes are applied to capture the physics on different temporal scales. Simulations start with an ITER 15MA baseline scenario profile generated by CORSICA (Kim et al 2015 Paper ITER_D_R9T8J9 v1.1 ). In BOUT ++ transport code, the plasma parameters ( n i , T i , T e ) and radial electric ( E r ) profiles are evolved to steady state. The initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients are calculated by inverting the plasma profiles inside the separatrix. SOL transport coefficients are assumed to be constants connected to the separatrix. A parametric scan for the anomalous thermal diffusivity ( χ i , χ e ) in the SOL is performed separately with E × B and magnetic drift included, and without any drift effects. The results show that when the diffusivity is smaller than a critical χ crit , the heat flux width λ q remains almost unchanged, which is roughly consistent with Goldston ’ s heuristic drift model (Goldston 2012 Nucl. Fusion 52 013009). Otherwise, it increases as a λ q ∝ χ 1 / 2 scaling resulting in a larger λ q . BOUT ++ six-field/two-fluid turbulence code is used to study pedestal and SOL turbulence dynamics and corresponding transport. In the turbulence simulation, pedestal is found to be peeling-ballooning unstable, which results in a larger λ q . Pedestal structure is also found to be important in determining the effective thermal diffusivity and could lead to changes in the divertor heat flux width.

An important task of the China Fusion Engineering Test Reactor physics design is to develop operation scenarios with high fusion power (1 GW), high bootstrap current fraction for steady-state and a plasma edge compatible with heat and particle exhaust. To achieve these goals, triangularity (delta) effects on the fusion performance of two candidate scenarios, with or without reversed magnetic shear (RS), namely conventional H-mode and RS H-mode, are evaluated using core-edge coupled integrated modeling in this paper. For fixed pedestal density, it is shown that higher delta is favorable for higher fusion performance in the conventional H-mode scenario while the fusion performance decreases with increasing delta in the RS H-mode scenario. In conventional H-mode, the higher fusion performance at high delta mainly comes from a higher pedestal temperature as predicted by EPED in combination with stiff core kinetic profiles. In the RS H-mode scenario with a local reversed shear region, the profiles are non-stiff and a strong internal transport barrier (ITB) exists at low delta. This results in higher density and temperature inside the ITB for low delta, leading to higher fusion power. If the pedestal temperature is kept fixed, in both scenarios the significant increase in pedestal density, which extends into the core, dominates at high delta and leads to much higher fusion power. For conventional H-mode, destabilization from increasing delta is partially balanced by stabilization due to increasing nu*. Since the normalized heat sources are quite similar, it results in minimal changes in the temperature profiles except for the lowest density case. For RS H-mode, destabilization from increasing delta is approximately balanced by stabilization due to increasing nu* in foot region, but a strong temperature ITB is still evident for low delta. The ability to take advantage of the high pedestal density in conventional H-mode and reversed shear scenario depends on its compatibility with edge density requirements from efficient heat and particle exhaust. Transport analysis is presented to elucidate the roles of delta, collisionality and magnetic shear in altering the profiles and the ITB, which contribute to the different behavior in the two scenarios.

The Chinese Fusion Engineering Test Reactor (CFETR) bridges the gap between ITER and a fusion power plant (FPP). The primary objectives of CFETR are: similar to 2 GW of fusion power, producing similar to 700 MW of net electric power, demonstrate tritium self-sufficiency, operate in steady-state and have a duty cycle of 30-50%. CFETR is in the pre-conceptual design phase and is currently envisaged to be a four-phase machine (from phase I P-fus similar to 200 MW to phase IV P-fus similar to 2 GW). In 2016 the EU and China began a collaboration on topics relating to nuclear fusion research and one topic of the work is on CFETR and DEMO. This contribution documents the progress on the collaboration on systems codes studies of CFETR. Systems codes attempt to model all aspects of a fusion power plant using simplified models (0-D, 1-D) and capture the interactions between plant systems. This allows the user to explore many reactor designs at a high level and optimise for different figures-of-merit (e.g. minimise major radius, R-0, or maximise fusion gain, Q). The EU systems code used for this work is PROCESS, which is the systems code used to create the EU-DEMO baseline designs. This paper details the work on analysing a 2018 CFETR design point in EU systems code PROCESS and the feasibility of the design with regards to meeting the performance objectives and operation of the machine. The work comments on the four-phased nature of the device and the systems code output focuses on phase IV. In combination with the systems code, an uncertainty quantification tool is used to investigate the sensitivity of a CFETR design point to changes in the input assumptions in the systems code. This paper details sensitivities of the CFETR design and shows that given the specified inputs and the uncertainties there are a reasonable number of feasible design points around the CFETR phase IV design point that still fulfil the high-level objectives of the machine.