Fusion energy is accelerating through conventional (DEMO) and alternative compact reactor designs, that are potentially faster and cheaper to build (e.g., ARC, STEP). Power exhaust is a key challenge and a potential show-stopper for all these designs. Recent experiment show the key benefits of strongly shaped Alternative Divertor Configurations (ADCs) [1-4], demonstrating their potential as a...
Consistent with the cost and complexity of the ITER full tungsten (W) actively cooled divertor, a comprehensive physics basis has been established over more than two decades (see [1] and references therein). This first-of-a-kind component, of unprecedented size and lifetime requirement, has now moved into the series production phase [2]. This does not mean, however, that physics studies in...
The stellarator’s steady-state capability offers inherent advantages for fusion power plants (FPP), including disruption-free operation and access to higher densities beyond the Greenwald density limit. However, reconciling particle exhaust and retention while fulfilling mandatory requirements of divertor life-time survival remains a critical challenge for reactor-relevant divertor operation...
The release of the new European DEMO LAR baseline affects the divertor design mainly in terms of input loads (e.g. heat flux) and poloidal profile. Two solutions are investigated, named “long leg”and “short leg” respectively. The long leg is defined and assumed as baseline, while the short leg is under further SOLPS analyses for complete investigations.
For both options, the grazing angle on...
Successful operation of fusion power plant (FPP) depends on a particle and power exhaust strategy which simultaneously facilitates good core performance. “Infinity Two” is Type One Energy’s proposed design for a practical FPP with a robust baseline physics solution and a conservative design margin [1]. It is four-field period, aspect ratio A = 10, quasiisodynamic stellarator with improved...
Optimizing divertor systems and controlling plasma exhaust are critical challenges for reactor-grade magnetic fusion devices such as ITER and future fusion power plants. Achieving these goals requires rapid, accurate modeling of boundary plasma behavior. Traditional high-fidelity edge plasma simulations, while scientifically valuable, are computationally intensive and were not designed for...
High heat loads in fusion reactors still remain as one of the most pressing problems to solve for the realisation of reliable power plants. To this end, divertor components are designed and developed to intercept and manage the large heat fluxes arising from the upstream plasma. Divertor diagnostic systems are utilised in experimental reactors to address the immediate challenges of real-time...
The TCV tokamak contributes to the development of nuclear fusion energy with proof-of-principle experiments and by validating models that are used to predict reactor performance. As part of the Swiss Roadmap for Research Infrastructures, the SPC is upgrading TCV to test a tightly baffled, long-legged divertor (TBLLD), a novel concept designed to enhance power exhaust capabilities with minimal...
The detached divertor regime has been demonstrated to be effective in mitigating steady-state particle and heat loads to divertor plasma-facing components in current devices and is a key aspect of operating regimes in future high power devices including ITER [1]. Alternative divertor configurations, such as the Super-X [2] offer advantages to conventional divertors such as a wider operating...
Negative Triangularity (NT) configurations exhibit higher energy confinement compared to the conventional Positive Triangularity (PT) configurations. Experiments on TCV [1] and DIII-D [2] have shown that NT L-Mode plasmas can achieve confinement comparable to H-mode, with $\beta_N$ up to 2.8 (2 in stationary state) demonstrated in TCV. This suggests the potential for high-confinement L-Mode...
The Tokamak Divertor serves as a critical component within fusion reactors, essential for managing plasma exhaust and ensuring the stability and efficiency of nuclear fusion devices. It effectively remove and contain particles such as helium ash, fuel impurities and heat-dissipating particles, thereby enhancing plasma stability and extending reactor lifespan. A Tungsten base alloy divertor...
Understanding the mechanisms that govern heat and particle transport in the divertor region is critical for the design and operation of future fusion reactors. Turbulent cross-field transport plays a key role in determining the heat flux distribution at divertor targets, affecting both the peak heat load and the overall power exhaust scenario. A key metric for characterizing heat flux...
The DIVertor GAs Simulator (DIVGAS), developed by the Vacuum group at the Karlsruhe Institute of Technology (KIT), offers a powerful and reliable framework for optimizing and evaluating divertor design – a critical component in advancing fusion technology in next-generation fusion reactors. The DIVGAS framework features two powerful modules – a deterministic one and a stochastic one – allowing...
In the recent experimental campaign OP2.2, the neutral gas pressures previously measured in the subdivertor of Wendelstein 7-X could be confirmed and improved with subdivertor neutral gas pressures of 3$\cdot10^{-3}$\,mbar routinely reached in standard as well as high iota configuration. Those two magnetic field configurations differ by the number and positions of the edge magnetic islands and...
A future nuclear fusion reactor demands its plasma-facing components (PFCs) to be able to handle the generated heat fluxes. For the divertor to survive continuous operation, mitigating the incoming heat loads is essential [1]. An established approach for reducing the heat loads is by injecting low to medium-Z impurities [2], which stimulates radiation emission in the plasma edge region....
Power exhaust remains a key challenge for tokamak-based nuclear fusion, requiring accurate prediction and control of heat loads on divertor targets. Strategies such as increasing divertor closure and exploring alternative divertor configurations (ADCs) are central to mitigating target heat and particle fluxes. The Tokamak à Configuration Variable (TCV) [1] is uniquely equipped to investigate...
Liquid metal (LM) has been conceptualized for use as plasma-facing component (PFC) in future fusion devices [1]. Being accessible to self-repairing and self-replenishment, thanks to the nature of the liquid phase, has been attractive for being applied in future long-run but less-maintenance fusion devices. This encourages the promotion of such research field throughout the years [2-5]. Despite...
UEDGE simulations of a “chimney” divertor, utilizing mid-leg pumping upstream of the divertor target along the outer baffle, predict the formation of a stable radiation front between the pumping plenum and X-point. The mid-leg pumping plenum is proposed as an engineering solution to stabilize the detachment front location downstream of the X-point, maintaining a hot X-point ($\rm T_{e,Xpt}$ ~...
The divertor, being the most heavily loaded component of a magnetic confinement fusion device, must withstand high heat flux (HHF) loads and intense neutron irradiation during fusion operation. Established designs for plasma-facing components (PFCs) in the divertor region comprise a combination of monolithic tungsten (W) armor blocks and a copper (Cu) alloy heat sink. One established design is...
The huge heat load onto divertor is a crucial issue in fusion reactor. While the radiative impurities are necessary for achieving divertor detachment especially for the future tokamak [1], it is also found to have essential effects on ELM control [2]. It implies the possibility of simultaneous control of the transient and steady-state heat load by impurity seeding. Therefore, it is necessary...
A large driver of future fusion reactor size is the need to handle transient events that could potentially cause re-attachment, which pushes the capabilities of conventional divertors [1]. Liquid metals are an attractive solution to transients due to vapor shielding [2] whereby the temperature of the plasma facing component (PFC) becomes clamped even at excessive plasma heat fluxes, such as...
Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m2), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its lifetime. Therefore, attention has to be paid to the thermal-hydraulic design of its cooling system. It is necessary to take effective cooling methods from the...
The exhaust of power as well as the He particles produced by the fusion reactions in a nuclear fusion reactor remains one of the key challenges. As a possible solution for this problem Alternative Divertor Configurations (ADCs) have been studied in many tokamaks worldwide, like TCV \citep{Reimerdes_2017,Theiler_2017}, DIII-D \citep{Soukhanovskii_2018}, NSTX \citep{Soukhanovskii_2016} and...
The X-point radiator (XPR) plasma regime displays favorable properties with regard to power exhaust in tokamaks. An H-mode-like confinement quality, a detached divertor, and the suppression of type-I ELMs are achieved simultaneously. XPR scenarios may also pave the way for more compact and cheaper divertor solutions, as demonstrated on ASDEX Upgrade [3]. The XPR regime was realized on...
It has now been demonstrated experimentally in several research tokamaks that a controlled X-point radiator (XPR) under H-mode conditions can not only provide for a fully detached divertor, but also yield a naturally more ELM-stable regime [1-3]. This is therefore a rather attractive scenario for reactors, especially those operating with tungsten (W) divertors, including ITER, since it is well...
An X-point radiator (XPR) features a stable, cold, and dense plasma surrounded by a highly radiative mantle above the X-point inside the confined region, providing a dissipated power fraction larger than 90%, fully detached divertor targets, and ELM mitigation, and is considered a potential solution for the power exhaust challenge in future fusion reactors. The XPR-like regime is observed in...
Experiments on the TCV tokamak have enabled the identification and detailed characterization of a novel X-point radiator regime, the X-point target radiator (XPTR) [1], which forms at a secondary X-point embedded along the outer divertor leg in the X-point target divertor geometry [2]. Unlike the conventional X-point radiator regime, the XPTR spatially decouples the radiator from the confined...
A substantial challenge regarding the realisation of magnetic confinement fusion (MCF) reactors is the reliable exhaust of power and particles. In this regard, plasma-facing components (PFCs) in the divertor region have to withstand high particle and heat flux loadings in combination with sustained fusion neutron irradiation. The latter inevitably leads to the deterioration of desired...
Actively water-cooled W/Cu divertor are considered as one of the primary choices of technical solution for the high heat flux removal in fusion reactor divertor. ASIPP initiated the first W/Cu divertor project in 2012, and then the EAST upper divertor was upgraded with W/Cu components in 2014 [1]. Afterwards, the W/Cu monoblock technology was further improved and passed the ITER qualification...
Fifty four Divertor Cassette Assemblies are to be manufactured to complete the toroidal ring of the full tungsten (W) ITER divertor. The Cassette Assembly consists of Outer Vertical Targets (OVT), Inner Vertical Targets (IVT), Dome and Cassette Body [1]. Three Domestic agencies (DAs) and their suppliers are involved in the procurement. The Japanese DA is in charge of OVT manufacturing, the...
Wendelstein 7-X, located at IPP in Greifswald, Germany, is the largest stellarator in the world with modular superconducting coils. It started plasma experiments with a water-cooled first wall including a carbon fiber reinforced carbon (CFC) based divertor in 2022, allowing for long pulse operation.
As a next step, plasma performance of a stellarator has to be demonstrated with carbon-free...
The Divertor Tokamak Test (DTT) facility is an experimental reactor under construction at ENEA (Frascati, Italy). The goal of the project is to demonstrate the feasibility of various divertor configurations and materials, identifying the most efficient in terms of power exhaust handling during fusion reactions. Some of the most demanding events during the lifespan of a high magnetic field and...
Recent progress in describing the H-mode operational space and access to small-ELM regimes via the separatrix operating space (SepOS) framework highlights potential pathways towards viable integrated scenarios in next step devices like ITER and SPARC. However, remaining challenges in extrapolating integrated scenarios due to uncertainties in power width and impurity concentration scalings...
The exhaust in a DEMO-class tokamak requires continuous operation in detachment [1]. In highly radiative detached regimes edge-localized modes (ELMs) may be suppressed, reducing reliance on RMP coils for ELM suppression. However, these regimes are close to radiative plasma limits which trigger disruptions that threaten machine-integrity. In stark contrast to conventional reactors, this...
The Wendelstein 7-X (W7-X) experiment has shown remarkable success in core plasma performance, achieving record triple product values in a stellarator geometry. This success has placed the stellarator as a top-contender in the race for fusion energy. However, heat and particle exhaust remains one of the outstanding topics to be addressed prior to the design of any future stellarator reactor....
The stellarator concept offers a promising pathway toward achieving nuclear fusion as a scalable, carbon-free energy source. Wendelstein 7-X (W7-X), an optimized stellarator experiment, aims to provide a proof-of-concept for this approach [1]. W7-X employs the island divertor concept as its plasma exhaust solution, utilizing a chain of magnetic islands at the plasma boundary, intersected with...
The plasma exhaust concept of the Wendelstein 7-X (W7-X) stellarator is based on the island divertor configuration, which exploits the interaction of magnetic islands with ten discrete carbon targets. These targets are designed to cope with multiple magnetic configurations featuring different island chains at the edge of the machine. They are therefore in an 'open' configuration with minimal...
For the long-pulse operation of the fusion rectors, the detachment phase is mandatory to protect the divertor from the overheating. The nitrogen-induced detachment has been widely applied but also been limited by its strong wall retention effect and the chemical activity. On EAST, under the ITER-like divertor, boron (B) coating and the long-pulse discharge conditions, N_{2} is firstly used to...
DIII-D is planning to install a new baffling and pumping structure in the upper divertor to test the concept of mid-leg pumping as a mechanism to passively stabilize the detachment front, maintaining a hot X-point (T$_{e,Xpt}$ $\sim$ T$_{e,sep,OMP}$) simultaneously with a detached divertor target (T$_{e,targ}$ < $\sim$5 eV). Adopting this innovative pumping concept has the potential to be a...
Managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge for fusion energy. Since this exhaust power fluctuates due to plasma disturbances, active power exhaust control is essential for reactors: a loss of detachment leads to target destruction while excessive cooling can trigger a highly damaging disruption. However, maintaining acceptable...
Control of particle recycling and neutral pressure in the divertor is a key challenge in sustaining high-performance, steady-state operation in fusion devices. This presentation highlights recent developments in the Large Helical Device (LHD), where a Closed Helical Divertor (CHD) with in-vessel pumping has been implemented to enhance neutral compression and control recycling.
The first part...
The tokamak divertor is subjected to huge heat load, including both the transient heat load due to the edge localized modes (ELMs) and steady-state heat load in between ELMs. Exploring the edge plasma solution compatible with the high-performance plasma is one of the key issues to achieve high-performance steady-state operation of magnetically-confined fusion reactors in the future.
Numerical...
Recent dedicated experiments combined with a new modeling suite have advanced the crucial topic of core edge integration and power exhaust for fusion plasmas substantially. We report on experimental findings with reactor relevant seeding gases that establish a core edge integrated boundary solution. A new core-edge integrated modeling framework has been validated on these experiments and is...
