A simple extrapolation of the ITER Q=10 H-mode divertor target heat-load specification assuming mitigated type-I ELMs is not sufficient for the exhaust concept required in EU-DEMO provided the increased power level of Pedge = 300MW entering the edge region. The combination of an anticipated reduced power fall-off length of only a few mm in the SOL at a plasma current Ip=20MA whilst keeping the...
The Wendelstein 7-X (W7-X) is an advanced stellarator device operated in Greifswald, Germany, to provide the proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high-performance, steady-state HELIAS operation. During the experimental programs, starting with the first operation phase OP1.1 (Dec. 2015 - March 2016) up to OP1.2a...
SPARC is a compact, high-field short pulse ICRF heated tokamak ($B_{0}=12.2$ T, $R_{0}=1.85$ m, $\tau_{flattop}=10$ s, $P_{rf}=25$ MW) with a close-fitting tungsten first wall designed to achieve its mission goal of $Q_{fus}>2$ with significant margin. Construction has begun at a new site in Devens, MA and operations are scheduled to begin in mid-2025. Although the baseline design relies on...
Significant advances have been made on the physics design of full tungsten divertor and the edge modeling for Chinese Fusion Engineering Testing Reactor (CFETR). CFETR is proposed by Chinese fusion community to bridge the gap between ITER and DEMO with fusion power up to GW level [1]. One of the key challenges is that the divertor solution for CFETR must meet requirements beyond that of ITER...
The STEP programme aims to demonstrate viability of a GW-scale spherical tokamak reactor, with parameters around: geometric axis 3.6m, aspect ratio 1.8, elongation 2.9, plasma current 20MA, on-axis magnetic field 3.2T, core radiated power 350MW, with steady state power crossing the separatrix 150MW. Here we give a physics overview of the studies driving forward the STEP divertor concept...
SONIC divertor code enables simultaneous calculations of seeding impurity (Ar) and fusion product (He ash) transport. He exhaust has been investigated in JA DEMO, where exhaust power ($P_{out}$ = 250 MW), ion flux ($\Gamma_{out}^{D}$ = $\rm 1x10^{22} s^{-1}$) and He ion flux ($\Gamma_{out}^{He}$ =$\rm 5.3x10^{20} s^{-1}$, corresponding to 1.5 GW fusion power) were given at the core-edge...
The new high field superconducting divertor tokamak test facility (DTT) [1] presently under construction is devoted to specifically study power exhaust solutions in regimes as close as possible to those foreseen in DEMO fusion reactor in terms of power crossing the separatrix, $P_{sep}/R$, and heat flux decay length, $\lambda_q$. The first DTT divertor will use the ITER-like technology based...
Safe plasma exhaust with high core performance is one of the remaining challenges on the development path towards a fusion reactor. Alternatives to conventional single-null divertors are assessed as risk mitigation. While multiple null divertors face severe technological challenges, long-legged divertors emerge as a promising option [1] where combination with tight baffling may further...
The realization of EU-DEMO will pose unprecedented challenges to the power exhaust system. Its plasma-facing components will have to cope with a steady-state load of about 450 MW (fusion + additional heating). It is expected that creating and controlling a proper radiating edge region will isotropically dissipate ~ 300 MW, with the remaining 150 MW flowing in the Scrape-off Layer towards the...
Next steps machines such as ITER and DEMO will face unprecedented challenges related to heat exhaust. Impurity seeding will play a key role in spreading power on a sufficiently large surface area. Simulating impurity radiation in a given scenario requires i) radiation functions for the different ion stages and ii) the spatial distribution of the ion densities. The latter results from the...
The kinetic simulation method has been employed to study the plasma-wall interaction mechanism in various plasma conditions: multi-component plasmas, electronegative plasmas, and dusty plasmas. The negative species are considered to enter the sheath region from presheath side with truncated Maxwellian distribution, and the ions satisfy the Bohm and Bohm-Chodura conditions. The sheath...
Characterizing the scrape-off layer (SOL) transport of Wendelstein 7-X (W7-X) is essential to assess the efficiency of its unique exhaust concept, the island divertor configuration. Insights into the SOL dynamics in attached and detached conditions can be gained by the measurement of particle flows. Investigations of impurity flow velocities and line-radiation have been carried out with the...
Divertor detachment operation is of critical importance for future long pulse high power tokomaks, such as ITER and CFETR[1]. Partially detached divertor conditions are foreseen for ITER and DEMO. In conventional standard divertor (SD) operation, the effect of drifts on SD detachment is very strong, and the outer target heat loading is very higher than inner target with favorable-Bt...
It is essential to control the plasma flow onto the divertor targets to avoid unacceptable heat load and erosion, and the contamination due to the sputtered impurities to avoid the degradation of the performance of confined plasma. Thus, comprehensive understanding of the plasma and impurity transport in the scrape-off layer (SOL) is necessary for exploring the divertor solution for the future...
The SPARC tokamak is a compact, high-field short pulse device (B0 = 12.2 T, R0 = 1.85 m, τflattop = 10 s) that plans to begin operations in mid-2025. It will execute a series of mission-driven campaigns to close science gaps to inform the design of the ARC fusion pilot plant to begin operation in the early 2030’s. Accomplishing this requires a versatile plasma diagnostic set for use in...
A self-consistent 2D model is presented for transport in boundary plasma and plasma-facing material walls. Plasma dynamics in the domain is represented by a 2D collisional plasma fluid model in the edge-plasma code UEDGE [1], and transport of hydrogen and heat in the wall material is represented by a system of 1D (into the wall) reaction-diffusion equations solved in the code FACE [2]. To...
We will present recent progress of alternative divertor geometry studies in TCV, focusing primarily on the effects of total flux expansion (Super-X), outer target poloidal flux expansion (X-Divertor) and additional X-points near the target (X-Point Target) or near the primary X-point (Snowflake-Minus-LFS). These geometrical features are combined with improved divertor closure using the new TCV...
SOLPS-ITER is used to investigate power sharing between the upper and lower outer divertors in a balanced double null magnetic configuration, revealing that asymmetric divertor conditions drive SOL power flows which can result in a large power sharing imbalance between divertors. Double null (DN) or near double null magnetic geometries are potential configurations for a future tokamak reactor...
A comprehensive analysis of the H fuel cycle, using experimental measurements aided by input from modeling, is presented for attached and detached plasmas. This analysis focuses on the particle transport processes in the Wendelstein 7-X island divertor is presented. This analysis allows an assessment of the status-quo and will quantify the optimization potential in particle collection,...
Optimization of divertor concepts for the 3D stellarator boundary is still only in its beginning stages. Multiple requirements must be satisfied for a properly functioning divertor, including optimized particle exhaust for density control, significant power exhaust to ensure the survival of plasma-facing components, and sufficient impurity retention in the scrape-off layer (SOL) to avoid undue...
The divertor is one of the key components of the EU-DEMO reactor. The development of a reliable solution for the power and impurity particle exhaust is recognized as a major challenge toward the realization of DEMO. The pre-conceptual design activities for the EU-DEMO divertor are carried on considering two project areas: the ‘Target development’, focusing on the design of the vertical targets...
The investigation on effect of the flowing liquid Li limiter (FLiLi) on fuel and impurity, and heat fluxe during high confinement mode (H-mode) plasma was recently performed in EAST, aiming to provide an alternative resolution for the divertor design of a future DEMO device. Four generations of liquid Li limiters have been have successively designed and tested including a thin flowing film...
The Divertor Tokamak Test facility (DTT) [1] is a fusion device currently under construction in ENEA Frascati. Its main scientific goal is to investigate advanced solutions for the heat exhaust in future fusion power plants, such as DEMO. A key step in the success of DTT is the development of integrated plasma scenarios where good core performance is achieved together with acceptable...
Experimental observations on TCV [1] and DIII-D [2] have shown that negative triangularity L-Mode discharges can exhibit H-mode grade confinement, opening the possibility for high confinement reactors that side-step the challenges associated with H-mode such as ELMs, narrow scrape-off layer widths, and density control. To ensure safe power exhaust that protects the plasma facing components,...
Injection of boron (B) and boron nitride (BN) in powder form into the upper closed divertor at DIII-D showed a substantial drop in divertor electron temperature from 30 eV to below 5 eV, increase in divertor neutral compression by up to an order of magnitude, and transition into stable detachment [1]. A decrease in wall fueling, main chamber neutral pressure, and the reduction of oxygen,...
Future large, high power fusion devices such as ITER and its successors will require strong dissipation of the plasma energy, momentum and particles reaching open field lines before reaching the surfaces of the divertor; detachment will play a dominant role in that dissipation. This contribution presents initial observations of divertor detachment from the MAST Upgrade spherical tokamak in...
Divertor recycling control is enabled by a new tungsten lower divertor on EAST, featured by a right-angled closed corner joining the vertical and horizontal target plates. ELM mitigation is observed as the outer strike point moves from the vertical target to the horizontal target with a significant reduction of the pedestal density gradient. Furthermore, the new closed corner divertor exhibits...
Divertor detachment with medium-Z impurities seeded through gas puffing can entail radiating regions within the last closed flux surface. The lithium vapor box divertor seeks to detach via near-target low-Z lithium evaporation with the result that such a radiating region does not form. We show SOLPS-ITER predictions for the effect of a lithium vapor box divertor on NSTX-U. Past work has shown...
Results from calculations of the neutron flux in the divertor cassette body for the Helium Cooled Pebble Bed and Water-Cooled Lithium Lead concepts are presented in this paper. For both cases under investigation, the same divertor model setup and designated DEMO neutron source were utilized. Neutron transport calculations were performed using the MCNP6 and the FENDL-3.1 nuclear data library....
Within the roadmap of the EU-DEMO reactor design, the divertor design is being developed in the framework of the EUROfusion Consortium Work package “Divertor”, sub-project DEMO (WPDIV-DEMO).
The ultimate objective of this sub-project is to deliver at least one holistic design concept and feasible technology options for the DEMO divertor and limiter eligible for the subsequent Engineering...
Time-dependent SOLPS-ITER simulations have been extensively applied to various problems such as actuator design, feedback control through system identification, and physical interpretation of dynamic problem. Most SOLPS-ITER simulations focus on steady-state solutions, and the EIRENE module, a neutral solver, couples in a time-independent mode assuming quasi-relaxation within time step of 1e-3...
Over the first phase of WEST exploitation, the full tungsten environment consisted in a mixture of coated tiles and actively cooled ITER-like monoblocks. Using a combination of LHCD and ICRH heating, diverted L-mode scenarios were extended up to 50 s and stationary heat loads reaching 6 MWm-2 were deposited on ITER-like monoblocks. Despite input power levels generally above expected thresholds...
Due to its unique property combination tungsten materials are the preferred choice for high-heat-flux-loaded areas in future fusion power plants. However, tungsten has a high brittle to ductile transition temperature and is prone to operational embrittlement due to high temperature and/or fast neutron irradiation. Tungsten fibre-reinforced tungsten composites utilize extrinsic mechanisms to...
Plasma-facing components based on tungsten fiber-reinforced tungsten composites
Y. Maoa,c, J.W. Coenena,e, J. Rieschb, X.Tanc, C.Chenc, A. Terraa, C.Liud, T. Höschenb, Y. Wuc, Ch. Broeckmannd, R.Neub,f, Ch. Linsmeiera
a Forschungszentrum Jülich GmbH,...
Divertor plasma-facing components (PFCs) of future fusion devices will have to deal with even more extreme heat loads than encountered at present. For ITER, it is expected that the current monoblock concept, where tungsten blocks are mounted on a copper cooling duct, will be sufficient. However, it is unsure whether these monoblocks can withstand the even more extreme conditions expected in...
Plasma edge codes are the workhorse for divertor design for future machines. Models that correctly account for kinetic neutrals, anomalous transport, and detailed wall geometry are essential to extrapolate current knowledge towards reactors. Yet, the required highly collisional regimes and large size lead to inacceptable runtimes. We report on the status and remaining challenges of some...
Heat and particle exhaust in tokamaks is determined by a complex interplay between plasma transport processes, including turbulence, and a set of physical phenomena due to the interaction between the plasma, the solid wall, recycling or injected neutrals and intrinsic or seeded impurities. The comprehensive modelling of the physics at play requires a consistent integration of each of these...
The control of a stable detachment solution that is compatible with both the core and edge may be crucial for the operation of reactor tokamaks in steady-state. In tokamaks there exist various methods of accessing and maintaining detachment, including fuelling, impurity seeding, and varying the heating power of the machine. How detachment is accessed and how its extent from the target...
A viable magnetic fusion power plant has to combine very high plasma density and temperature in the core region, in order to maximize fusion reactions, with cold plasma conditions in the peripheral region compatible with long life expectancy of plasma-facing components. In this contribution, taking inspiration from recent work on DIIID tokamak (see ref. 1), we examine this crucial issue for...
The X-point radiator (XPR) is an attractive scenario to solve the power exhaust problem in future fusion devices. In ASDEX Upgrade (AUG), experiments with an XPR showed a dissipated power fraction larger than 90 %, fully detached divertor targets and ELM suppression with a moderate confinement degradation [1]. Recently, a reduced model [2] was derived to explain the physical mechanisms for...
Divertor and exhaust scenario design for future reactors such as ITER and DEMO heavily rely on plasma edge codes as SOLPS-ITER [1]. In practice, the design process often proceeds through large parameter scans to explore the operational space, whereby divertor shape, magnetic field and model parameters are manually tuned to improve the performance of the design and meet various physics,...
EU-DEMO reactor operation is expected to go beyond the ITER requirements of a semi-detached divertor regime [1], and will have to achieve the even higher levels of dissipation needed to demonstrate a realistic power exhaust solution for a fusion power plant. Sustaining the desired degree of detachment will require reliable real time (RT) control. In addition to the actuators (e.g. the impurity...
While detachment is in present-day machines routinely achieved under different conditions during the flattop of a discharge by either density ramps or extrinsic impurity seeding, it has not yet been shown to be accessed and controlled throughout the full duration of a discharge. For future machines, it is essential to be in detachment already during the power ramp up, the L-H transition and...
Significant progress has been made on the new lower tungsten (W) divertor with closed geometry and active water-cooling capability for steady state operations in EAST since 2021. The latest experimental results demonstrate that the new divertor exhibits strong particle exhaust capability and relatively high neutral retention in the divertor region, which facilitate both impurity screening and...
The edge-localized-modes (ELMs) in magnetic fusion reactor raise a major concern not only due to the degraded core confinement, but also due to the adverse effects against plasma facing components [1]. In the existing medium size tokamaks, the application of resonant magnetic perturbation (RMP) has proven effective in suppressing ELMs without being limited to ELM-crash-mitigation. Nonetheless,...
One of the most promising approaches to tackle the power exhaust problem in a divertor tokamak is the so called X-point radiator (XPR). By the controlled injection of impurities into the plasma a radiation cloud localized in the vicinity of the X-point is formed. It was shown that up to 95% of the power absorbed in the plasma can be dissipated before reaching the divertor targets [1]. Recent...
