H-mode Operation Research Articles

First access to ELM-free negative triangularity at low aspect ratio

Abstract A plasma scenario with negative triangularity (NT) shaping is achieved on MAST-U for the first time. While edge localized modes (ELMs) are eventually suppressed as the triangularity is decreased below δ ≲ − 0.06 , an extended period of H-mode operation with Type-III ELMs is sustained at less negative δ even through access to the second stability region for ideal ballooning modes is closed. This documents a qualitative difference from the ELM-free access conditions documented in NT scenarios on conventional aspect ratio machines. The electron temperature at the pedestal top drops across the transition to ELM-free operation, but a steady rise in core temperature as δ is decreased allows for similar normalized β in the ELM-free NT and H-mode positive triangularity shapes.

Overview of the early campaign diagnostics for the SPARC tokamak (invited).

The SPARC tokamak is a high-field, Bt0 ∼12T, medium-sized, R0 = 1.85m, tokamak that is presently under construction in Devens, MA, led by Commonwealth Fusion Systems. It will be used to de-risk the high-field tokamak path to a fusion power plant and demonstrate the commercial viability of fusion energy. SPARC's first campaign plan is to achieve Qfus > 1 using an ICRF-heated, <10MW, high current, Ip ∼ 8.5 MA, L-mode fueled by D-T gas injection, and its second campaign will investigate H-mode operations in D-D. To facilitate plasma control and scientific learning, a targeted set of ∼50 plasma diagnostics are being designed and built for operation during these campaigns. While nearly all diagnostics are based on established techniques, the pace of deployment, relative to the first plasma, and the harshness of the thermal, electromagnetic, and radiation environment are unprecedented for medium-sized tokamaks. An overview of the SPARC diagnostic set is given, providing context to further details communicated by the SPARC team in companion publications that are system-specific. The system engineering philosophy for SPARC diagnostics is outlined, and the design and engineering verification process for components inside and outside the primary vacuum boundary are described. Diagnostics are mounted directly to the vacuum vessel as well as housed within a series of eight midplane and 24 off-midplane replaceable port plugs. With limited exceptions, signal conditioning, digitization electronics and cameras as well as lasers and microwave sources are localized to a series of five Diagnostic Lab spaces, totaling ∼350m2, located >15m from the center of the tokamak, on the other side of a 2.4m concrete shielding wall. A series of 31 large-scale penetrations have been included in the SPARC Tokamak Hall to facilitate integration of early campaign diagnostics and to provide upgradability.

Dependence of ELM instability on separatrix density in EAST long-pulse H-mode plasmas

Abstract The transition from small edge-localized modes (ELMs) to large ELMs has been repetitively observed in minute-scale long-pulse high-confinement mode (H-mode) discharges in 2017 EAST campaign. The appearance of large ELMs is found to be strongly correlated with the decrease in separatrix density due to the gradual decrease in fuel recycling during the long-pulse H-mode operations (LPHOs). By numerical scan of separatrix density with a fixed temperature profile, it has been found that the dependence of ELM instability on separatrix density is related to the competition between the ion diamagnetic stabilizing effect and destabilizing effect of pressure gradient and current density in the pedestal region, shedding light on a comprehensive understanding of different roles of separatrix density in ELM instability observed in EAST experiments. With a high separatrix density, the ideal ballooning mode could be destabilized near the separatrix, which is thought to help achieve small ELMs in EAST LPHOs. In 2021 EAST campaign, the experiment of large ELMs control has been performed through actively changing fuel recycling by moving strike point location (SPL) on the lower tungsten divertor target plate. It has been demonstrated that the mitigation of large ELMs is strongly correlated with the significant increase in separatrix density, which is thought to be attributed to a higher ionization source in the scrape-off layer (SOL) region by SOLPS-ITER simulation. The high ionization source in the SOL region is believed to provide a strong fueling effect near the separatrix and thus raise the local density, which is considered as an important reason for triggering ballooning instabilities near the separatrix and achieving small ELMs.

Helium plasma operations on ASDEX Upgrade and JET in support of the non-nuclear phases of ITER

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

Nonlinear simulations of the peeling-ballooning instability of super H-modes in the HL-3 tokamak

As the newly built tokamak in China, HL-3 will explore high-performance operation scenarios, such as super H-mode. The energy confinement and core parameters in the super H-mode can be much larger than that in the normal H-mode. Based on the pedestal simulation code EPED, the operation space of the super H-mode is obtained in HL-3. Magnetic shear decreases with increasing triangularity; consequently, a super H-mode can be achieved. The threshold of triangularity for accessing a super H-mode in HL-3 is around 0.4. By using BOUT++, a nonlinear simulation study of the pedestal instabilities in the super H-mode equilibrium is executed for the first time. As expected, the low n peeling mode, which can cause much of the energy loss (17%) from the pedestal region, is dominant in the super H-mode. Such a large collapse in the pedestal region would lead to a transition from super H-mode to H-mode. It is crucial to expand the parameter space of the super H-mode or mitigate the edge-localized mode (ELM) size for sustaining the super H-mode operations. The E × B velocity shear is found to play an important role in controlling the ELMs in HL-3. The small E × B velocity shear leads to a large growth rate but results in a small ELM size around the peeling boundary. The ELM size is closely related to both the growth rate of peeling-ballooning mode and the duration time of the linear phase. In contrast, a large E × B velocity shear can stabilize the instabilities near the ballooning boundary. Next, the parameter space of the super H-mode can be enlarged.

Power and isotope effects in the ITER baseline scenario with tungsten and tungsten-equivalent radiators in DIII-D

Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q 95 ∼ 3 and P IN/P LH ∼ 1–2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation (P LH) was determined experimentally without added impurities and found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20%–25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations C W = 10−5. Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at C W = 1 × 10−5, moreover Q = 8–10 would still be achieved for C W up to 3 × 10−5. In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities, indicating that hydrogen H-modes in ITER may not inform the D-T phase with respect to W accumulation and discharge survival. Compared to deuterium, the pedestal pressure in hydrogen is ∼25% lower, with much higher ELM frequency of 150 Hz, decreasing with input power. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2–3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3–0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous International Tokamak Physics Activity database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H 98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation. The new DIII-D results support a more attractive option for the ITER Research Plan with a short hydrogen phase for system commissioning, transitioning to deuterium operations as soon as possible to provide relevant conditions for deuterium-tritium operations.

Core performance predictions in projected SPARC first-campaign plasmas with nonlinear CGYRO

This work characterizes the core transport physics of SPARC early-campaign plasmas using the PORTALS-CGYRO framework. Empirical modeling of SPARC plasmas with L-mode confinement indicates an ample window of breakeven (Q &amp;gt; 1) without the need of H-mode operation. Extensive modeling of multi-channel (electron energy, ion energy, and electron particle) flux-matched conditions with the nonlinear CGYRO code for turbulent transport coupled to the macroscopic plasma evolution using PORTALS reveals that the maximum fusion performance to be attained will be highly dependent on the near-edge pressure. Stiff core transport conditions are found, particularly when fusion gain approaches unity, and predicted density peaking is found to be in line with empirical databases of particle source-free H-modes. Impurity optimization is identified as a potential avenue to increase fusion performance while enabling core-edge integration. Extensive validation of the quasilinear TGLF model builds confidence in reduced-model predictions. The implications of projecting L-mode performance to high-performance and burning-plasma devices is discussed, together with the importance of predicting edge conditions.

Observation of fast electron redistribution during saturated kink mode in high βp H-mode discharge with central heating in EAST tokamak

In the EAST tokamak, we have developed an internal transport barrier (ITB) high-confinement mode (H-mode) scenario characterized by dominant electron heating and centrally peaked electron temperature profiles, facilitated primarily through the combustion of lower hybrid current drive and electron cyclotron radio heating (ECRH). Hard x-ray diagnostics reveal a marked increase in the population of fast electrons with energy from 30 keV to 80 keV, concurrent with augmented ECRH power during H-mode plasma operations. This surge in fast electron population precedes the formation of the electron temperature ITB (Te-ITB). Within the Te-ITB H-mode discharge, a mild and long-lived m/n = 1/1 mode (where m and n denote the toroidal and poloidal mode numbers, respectively) emerges proximal to the ITB region. This mode precipitates a redistribution of fast electrons, contributing to an increase in the safety factor near the magnetic axis and thereby promoting the stability of the Te-ITB. Furthermore, we explore the influence of fast electrons on plasma pressure and examine the effects of the profile of fast electrons on the central Te. Strategies to maintain the m/n = 1/1 mode at a moderate amplitude are also discussed, highlighting their significance in the sustained management of Te-ITB.

Design and analysis of a PAM launcher at 4.6 GHz for a new LHCD system on EAST

To improve the Current Drive (CD) capability in long-pulse (up to ∼1000 s) H-mode operation, it has been decided to develop a new Lower Hybrid Current Drive system at 4.6 GHz with an active cooling Passive Active Multijunction (PAM) launcher on EAST. In this paper, both the radio frequency (RF) and the physical properties of this PAM are studied numerically. The same nominal parallel refractive index (N || = k ||c/ω, where k || is the parallel wavenumber, c the velocity of light, and ω the wave angular frequency) of 2.04 as the existing 4.6 GHz Full Active Multijunction (FAM) is chosen. Ray-tracing calculations indicate that good accessibility could be achieved when the LH waves radiate with this nominal N || in typical long-pulse H-mode plasmas. The coupling performance in terms of power reflection coefficient (R C), power spectrum, maximum electric field, power directivity (D P) and global CD capability is evaluated with the ALOHA code based on the linear coupling theory. Good coupling performance with averaged R C ⩽ 1% and D P ∼ 70% could be expected with the density (n e) in front of the PAM close to the cut-off value (n e_co). The simulated R C remains below 6.5% over a wide density range 0.5 ⩽ n e/n e_co ⩽ 10, which is similar to the plasma edge conditions produced by Edge Localized Mode activity. A detailed comparison with the existing 4.6 GHz FAM launcher is also performed.

First results of high density H-mode operation in metal-wall EAST tokamak

In metal-wall EAST superconducting tokamak, H-mode operation with plasma density close to the Greenwald density limit nGW has been achieved with radio frequency and NBI heating for the first time. Both gas puffing from horizontal plane and HFS pellet fueling were used for density ramp-up during the experiment. The confinement of H-mode gradually deteriorates with plasma density increasing. And the H-L transition can be observed after heating power dropping or pellet injection. In the density range of (0.6–1)×nGW, the divertor detachment occurs and causes an obvious confinement degradation. The maximum accessible density ne, max in H-mode phase deviates from the Greenwald scaling. It has been observed that the fraction ne, max/nGW is almost independent of the total heating power, but a high heating power is helpful to extend the duration of high density H-mode. And ne,max/nGW has an increase relation with the safety factor q95 varied by changing plasma current. Besides, it is also found that the discharges with low plasma current have a higher ne, max/nGW than those with high plasma current. All these findings will provide a good reference to the high density plasma operation for future metal-wall fusion devices.

Performance of a liquid Sn divertor target during ASDEX upgrade L-mode and H-mode operation

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

The power dependence of the maximum achievable H-mode and (disruptive) L-mode separatrix density in ASDEX Upgrade

In future fusion reactors disruptions must be avoided at all costs. Disruptions due to the density limit (DL) are typically described by the power-independent Greenwald scaling. Recently, a power dependence of the disruptive DL was predicted by several authors (Zanca et al 2019 Nucl. Fusion 59 126011; Giacomin et al 2022 Phys. Rev. Lett. 128 185003; Singh and Diamond 2022 Plasma Phys. Control. Fusion 64 084004; Stroth et al 2022 Nucl. Fusion 62 076008; Brown and Goldston 2021 Nucl. Mater. Energy 27 101002). It is investigated whether this increases the operational range of the tokamak. Increasing the heating power in the L-mode can induce an L-H transition, and therefore a power-dependent DL and the L-H transition cannot be considered independently. The different models are tested on a data base for separatrix parameters at the separatrix of ASDEX Upgrade and compared with the concept (SepOS) presented in Eich and Manz (2021 Nucl. Fusion 61 086017). The disruptive separatrix density scales with the power in good agreement to all models. Also the back transition from high to low (H-L) confinement shows an approximately Greenwald scaling with an additional power dependence according to the SepOS concept. For future devices operating at much higher heating power such a power scaling may allow operation at much higher separatrix densities than are common in H-mode operation. Preconditions to extrapolation for future devices are discussed.

Long-pulse H-mode operation with stored-energy monitoring for detachment feedback control with a new lower tungsten divertor in EAST

One of the key challenges for future fusion research is to mitigate the high steady-state heat load on the divertor target plates, and divertor detachment offers a promising solution. The Experimental Advanced Superconducting Tokamak (EAST) has developed several feedback control methods for divertor detachment. However, when an off-normal event momentarily disturbs the main plasma, impurity seeding may still be conducted by these methods for detachment, which probably drives the main plasma further away from its stable equilibrium or even causes the disruption. These off-normal events include excessive impurity seeding, loss of heating and dust droplets, which are not rare in current tokamak experiments, especially in long-pulse operation. To compensate for the drawbacks of these methods, we propose and develop a module of stored-energy monitoring to ensure stable plasmas in long-pulse operation. The stored energy usually decreases when the main plasma is away from its stable equilibrium, which is suitable for monitoring the state of the main plasma. Once the stored energy falls below a certain threshold, the module actively switches off the impurity seeding system. Without impurity seeding, the main plasma can recover with the increase in the stored energy. Only when the stored energy exceeds another threshold does the module switch on the impurity seeding to continue the detachment operation. The module function has been verified during the EAST radiative divertor experiments in the newly-upgraded lower tungsten divertor. A typical ∼20 s discharge in grassy-ELM H-mode regime with ∼5 MW source heating power is demonstrated with divertor partial detachment and good energy confinement by active impurity seeding (50% neon, 50% D2). The energy confinement factor is maintained at a high level, i.e. The electron temperature in the core region only has a slight change after the impurity seeding, while the electron density has a ∼10% increase. Furthermore, the ion temperature near the axis also has a remarkable increase. These achievements provide an important demonstration of the actively controlled radiative divertor mitigating the heat loads with good core confinement, which is an essential step toward steady-state operation of fusion reactors.

Experimental study of density gradient-driven micro-instabilities and the confinement degradation during H-mode in EAST

A broadband (BB) mode is observed by collective Thomson scattering diagnostics in repeatable shots of EAST and analyzed for the first time. This BB mode usually grows during L–H transitions, featuring a BB quasi-coherent mode with increasing frequency. During H-mode operations, it is characterized by steady-state BB in the high-frequency range (f ∼ 200–2000 kHz), at the electron scale (k θ ρ s = 1‒2), mainly driven by the density gradient, and is sensitive to the value of η e in the region of interest (ρ = 0.4‒0.8), where is the ratio of the normalized electron temperature gradient and density gradient, and the region ρ = 0.4‒0.8 usually has a relatively low collisionality (v eff < 5). The frequency of BB is found to be dependent on the electron temperature and density gradient, which is a typical feature of electron-driven turbulence. A negative correlation between the energy confinement and the intensity of the BB turbulence during H-mode has been found, which indicates a strong electron thermal transport induced by the BB turbulence. The BB significantly decreases the electron temperature and causes flatter electron temperature profiles in the region of interest (ρ = 0.4‒0.8), thus making η e decrease and the BB destabilize further. These characteristics of BB are related to the theoretical density gradient-driven trapped electron mode. It should be noted that this mode is not observed by other diagnostics in EAST, and shows very different features to the coherent modes in the edge.

Analysis of edge transport in L-mode negative triangularity TCV discharges

One of the major problems for future tokamak devices are ELMs (Edge Localized Modes) as they can lead to large, uncontrolled heat fluxes at the machine targets. For this reason, different techniques and alternative magnetic configurations are under study to mitigate or avoid these phenomena. One of the most promising among these studies is the Negative Triangularity (NT) configuration, which exhibits a global confinement comparable with H-Mode operation and, staying in L-mode, could enable an easier power exhaust dissipation due to a possible bigger heat flux decay length with respect to the conventional Positive triangularity (PT) H-mode. In this work, the fluid code SOLEDGE2D-EIRENE is used to study edge transport. Studies are made on discharges performed in the TCV device (Tokamak à configuration variable), which can create a variety of different plasma geometries thanks to 16 independently powered poloidal field coils and its open vacuum vessel. In order to understand if and how power and particle exhaust in NT differs with respect to those of the PT shape, four discharges in single null magnetic divertor configuration with fixed lower triangularity (δbot=+0.5) but with different upper triangularity (from δup=-0.28 to δup=+0.45) have been modelled. All of them are ohmically heated, L-mode deuterium plasmas, and in the high recycling regime. Moreover, these discharges were previously used in [4] to measure the heat flux decay length by IRT (Infrared Thermography), allowing us to make comparisons with modelling results.

Numerical analysis of divertor target heat load of I-modes in comparison with H-modes of EAST

Due to substantial edge transport of particles, I-mode operations offer a high potential for divertor heat load mitigation. In this work, the divertor parameters of I-mode operations on EAST have been investigated using the SOLPS-ITER code, and comparison with H-mode operations has also been made by modeling the I- and H-mode operation processes on EAST. The simulation shows that, for the same separatrix electron density at the outer midplane (OMP) , the upstream electron density of the I-mode is higher than that of the H-mode with no density pedestal, while the upstream temperature of the I-mode is almost the same as that of the H-mode with a temperature pedestal similar to that of the H-mode. As a combined result, the temperature and energy flux peaks of the I-mode are thus lower than those of the H-mode at the divertor target. Further parameter scanning investigation reveals that, under low-density conditions (), the peak energy flow at the target is reduced by ∼34.1%, in the case of the I-mode as compared to the H-mode, while the peak target temperature is dropped by ∼54.6%. Under high-density conditions (), on the other hand, the energy flux and temperature peaks are weakened by ∼28% and ∼30.1%, respectively. The upstream density at detachment onset of an I-mode is also lower than that of an H-mode, by 18.5%. These results suggest that I-mode operation is more appropriate for divertor heat load mitigation than H-mode operation.

Nonlinear evolution characteristics of peeling-ballooning mode under negative triangularity

&lt;sec&gt;Experiments on TCV tokamak have achieved high confinement mode (H-mode) operation with negative triangularity, and this mode shows quite different characteristics from those with the positive triangularity in experiment and simulation. Linear simulations for kinetic ballooning mode and peeling-ballooning(PB) mode without diamagnetic effect show that negative triangularity can enhance the instability of the ballooning mode and close access to the second stable region. However, the understanding of ELM for negative triangularity is not sufficient. Therefore, it is necessary to carry out further research on ELM with negative triangularity.&lt;/sec&gt;&lt;sec&gt;In this work, based on a series of equilibria with different triangularities in Tokamak, the nonlinear characteristics of negative triangularity of PB mode is investigated. It is found that the negative triangularity can destabilize the PB mode by a larger unfavorable curvature region, which will reduce the instability threshold, and thus limiting the increase of pedestal height. In the nonlinear phase, the pressure perturbation intensity with negative triangularity will extend to the top area and the bottom area in the low field side and bring about an earlier ELM collapse. Meanwhile, modes with different toroidal mode numbers are more likely to be triggered off and then grow and replaces the initial unstable mode, showing more obvious turbulent transport characteristics, which can play a role in the ELM energy loss.&lt;/sec&gt;

Overview of initial negative triangularity plasma studies on the ASDEX Upgrade tokamak

An overview of results of negative triangularity (NT) studies from the ASDEX Upgrade tokamak (AUG) is given. Moderate values of the triangularity in the range of δ average ≈ −0.2 have been obtained. Results from both unfavorable and favorable magnetic configuration in terms of high-confinement mode (H-mode) access are presented. In unfavorable configuration, the ion grad B drift points away from the active x-point, which usually results in a factor of 2 higher power thresholds to enter H-mode. In unfavorable configuration, low-confinement mode (L-mode) operation is confirmed at high auxiliary heating input powers MW at low collisionality. Under these conditions, plasmas with positive triangularity (PT) are usually in H-mode. Highest plasma energies have been obtained using a combination of neutral beam injection, electron- and ion cyclotron resonance heating, where values of the plasma beta of β N ≈ 1.7 have been reached in L-mode. While discharges with mixed auxiliary heating power input show moderate confinement on L-mode levels, plasmas heated with electron cyclotron resonance heating show H-mode like confinement. Operation in favorable configuration, however, leads to H-mode with type-I edge localized modes (ELMs). The power required for the L–H transition is comparable to the typical L–H power threshold in PT cases ( MW), and good confinement of typical PT H-modes is obtained. The ELM frequency is demonstrated to reach 700 Hz. These results show a clear effect of the NT configuration on the ELM behavior in H-modes, and that on AUG, unfavorable configuration is an essential ingredient if H-mode operation is to be avoided in NT plasmas. So far, in all AUG NT plasmas, power degradation is observed. Initial gyrokinetic simulations of the edge plasma in unfavorable configuration reveal the ion-temperature gradient mode as the fastest growing mode with a subdominant trapped-electron-mode.

Design of the Cooling Water System for the EAST Lower Divertor

The new lower tungsten divertor is a crucial component to achieve the targets of long pulse and high <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\beta $ </tex-math></inline-formula> H-mode operation of Experimental Advanced Superconducting Tokamak (EAST) device. The cooling water system (CWS) of lower divertor should have sufficient heat removal capability and be able to cope with pressure transients caused by the potential water hammer events, which makes the design challenging. In this study, the system process and details of main component are determined based on the design requirements. The thermal-hydraulic analysis model has been developed by Flowmaster (FLOWMASTER) to conduct the heat transfer analysis. Furthermore, the system water hammer transients under one pump trip condition have been analyzed. The results show that the flow distribution of each client is uniform, and the pressure and temperature of key points meet the design requirements. The maximum pressure fluctuation, 0.16 MPa, caused by the sudden closing of the check valve is within the acceptable range. The experiment data are in good agreement with the simulation values, which proves the correctness of the model and the feasibility of the design.

Design of a multi-configurations divertor for the DTT facility

The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation. The new high field superconducting divertor tokamak test facility (DTT) [1] is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs. Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered. The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume. In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.