Fifth Technical Meeting on Divertor Concepts

28 Oct 2025, 08:30 → 31 Oct 2025, 17:50 Europe/Vienna

Press Room (M-Building) (IAEA Headquarters)

KEY DEADLINES

21 June 2025 Deadline for submission of abstracts through IAEA-INDICO for regular contributions. Deadline for participants to submit their application via the InTouch+ platform

30 July 2025 Deadline for participants to submit their application via the InTouch+ platform

25 July 2025 Notification of acceptance of abstracts and of assigned awards

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Located at the very bottom of a magnetic fusion device in most designs, where impurities such as helium ‘ash’ are diverted, the divertor acts as the ‘exhaust pipe’ of the fusion machine and is where any excessive heat is channelled to. This configuration helps to produce ‘purer’ plasmas with better energy confinement — a critical parameter for the performance of a fusion device — ensuring the plasma is hot enough for long enough so that sustained fusion reactions can take place.

In ITER, the divertor will be made up of 54 ‘cassettes’, each weighing 10 tonnes. The conditions placed on the cassettes will be very demanding; facing steady heat fluxes of 10 to 20 megawatts per square metre, with parts exposed to temperatures of between 1000°C and 2000°C, the cassettes will need to be replaced by remote handling at least once during the machine’s lifetime. To deal with the extreme heat and damaging particles, the components facing the plasma will be armoured with tungsten, a material that has both low tritium absorption and the highest melting temperature of any natural element. Although ITER’s divertor design reflects the state of the art of our current understanding and capabilities from a physics and technology point of view, further developments will be required for future fusion power plants.

Objectives

The event aims to provide a forum for discussion and analysis of the latest findings and open issues related to divertors in fusion devices in the context of ITER, demonstration fusion power plants and next-step facilities. The participating authors are invited to put their abstract into this context and thereby provide contributions to this meeting that serve as a basis for discussions.

Target Audience

The event aims to bring together junior and senior scientific fusion project leaders, plasma physicists, including theoreticians and experimentalists, and experts (researchers and engineers) in the physics and technology of the divertor.

Guidance for Presentations

Presenters are recommended to contextualise their results in respect to broader applicability across different types and sizes of diverted magnetic confinement fusion devices. Emphasis should be given to highlighting relevance to DEMO/plant scale devices and DEMO/plant relevant operating scenarios. Scenarios such as long-pulse/steady-state operation, impact of transients/fluctuations, and the relevant timescales for the stability of a scenario (steady-state or time dependent). 

Additionally, it would support the meeting if presenters could share identified gaps in the risks being addressed by the divertor community. As well as lessons learned from the upgrades and modifications of devices. Modifications such as JET carbon to beryllium-tungsten wall change, DIII-D small angle slot divertor, and the new KSTAR tungsten divertor.

2025 IAEA TM on Divertor Concepts.pdf

Tuesday 28 October

09:30 → 10:00 Introduction and Welcome

Conveners: James Harrison (United Kingdom Atomic Energy Authority), Mr Matteo Barbarino (International Atomic Energy Agency)

10:00 → 11:15 Divertors for Next-Generation Devices

Convener: Anthony Leonard (USA)

10:00 Topical Session 1 - Divertors for Next-Generation Devices

Speaker: Anthony Leonard (USA)

10:05 Divertor shaping as a continuum strategy to solving the power exhaust challenge

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 power exhaust solution. Integration of ADCs in a reactor is complex: a compromise between power exhaust benefits and engineering feasibility is required [5]. Our experiments show that a continuum of ADC optimisation strategies exists: modest, yet strategic, divertor shaping can greatly enhance power exhaust performance [3] and control [4].

To study this continuum the Super-X, Conventional and an intermediate Elongated Divertor geometry (see figure [3]) were compared on MAST Upgrade. The Elongated Divertor has only half the total flux expansion increase of MAST-U’s Super-X divertor: lower than that of STEP [7] and comparable to that of SPARC [8], ARC [9] and the DEMO Super-X Divertor [6].

Crucially, the key benefits of the MAST-U Super-X Divertor over the Conventional Divertor are maintained in the moderately shaped Elongated Divertor.

1. Target heat loads are reduced beyond geometric spreading expectations.
2. The sensitivity of detachment is drastically reduced, improving real-time power exhaust control [4].
3. Reducing the detachment onset without any adverse core impact increases the operational window of the detached regime. This improved core-edge compatibility can enable reactor operation at lower core power losses [6] and/or lower upstream densities/fuelling (relevant for ELM-free scenarios and fuel cycle limitations [7]) and/or lower impurity concentrations.

Studying the physics driving these ADC benefits provides lessons on both divertor design – relevant for reactors - and exhaust physics/simulations and control [4] – relevant for ITER. This shows synergies between neutral baffling, poloidal leg length and total flux expansion enable power exhaust benefits from strategic divertor shaping; consistent with reduced and full models. This demonstrates ADCs offer a continuum of solutions, from modest to advanced, enabling balancing engineering complexity and reactor power exhaust performance, that can be tuned to an individual reactor’s specification.

[1 ] K. Lee. PRL. 2025. C. Theiler, this meeting. [2] D. Moulton. NF, 2024. [3] K. Verhaegh. Comm. Phys. 2025. [4] B. Kool. Nat. Energy, 2025, in press. This meeting. [5] R. Kembleton. FED 2022. [6] Xiang. 2021, NF; [7] Henderson. NF 2025; [8] Kuang, et al. JPP 2020; [9] Wigram, NF, 2019.

Speaker: Prof. Kevin Verhaegh (Eindhoven University of Technology)

10:45 Recent advances in the ITER divertor physics basis

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 support of the ITER divertor and its operation no longer take place. Rather, exploration has continued, from the point of view of both stationary and transient operation. It must also now focus on aspects related to the new ITER Research Plan (IRP) which accompanies the 2024 re-baselining activity [3,4]. The purpose of the present submission is to summarise some of the key new analyses undertaken since the physics basis was last summarised in the 3rd meeting of this IAEA TM Divertor Concepts series held in 2019.
As the component which intercepts >90% of the thermal plasma power exhausted into the scrape-off layer (SOL), the principal design driver for the ITER divertor has always been stationary power loading. Evidently, burning plasmas (PSOL ≥ 100 MW) represent the most severe test, both in terms of power density and fluence, but it is also important to assess the loading to be expected in the earlier, lower power phases of ITER operation, specifically the Start of Research Operations (SRO) campaign in the new baseline.
This SRO phase contains a short deuterium (DD) H-mode campaign, a primary objective of which is to test edge localized mode (ELM) control schemes in advance of the DT phase. This has motivated a detailed EMC3-Eirene study of divertor conditions and detachment access in the presence of resonant magnetic perturbations (RMP), using MARS-F plasma response modeling to provide realistic magnetic field geometries resulting from RMP coil phasings which maximise the X-point displacement figure of merit known to be correlated with ELM suppression in current devices [5]. Scenarios at n = 4 are found to be preferred over n = 3 or hybrid n = 3,4 perturbations, yielding smaller magnetic footprint width on the target and with the helical lobe structures situated closer to the original, unperturbed strike point, facilitating momentum and radiative losses (i.e. detachment).
For non-perturbed configurations, new, drift activated SOLPS-ITER studies have been dedicated to the assessment of performance during the DD H-mode phase, demonstrating that at the maximum expected PSOL ~ 40 MW, impurity seeding is likely to be required for both power load and upstream separatrix density control [6]. Efforts have also been directed towards new burning plasma simulations deploying a higher fidelity model of the divertor structure, permitting a more realistic accounting of neutral bypass conductances [7]. This generates neutral recirculation patterns absent in the existing SOLPS database, reducing far SOL target plasma temperatures and neutral pressures at the pumping duct, with implications for W target erosion, He exhaust and upstream He separatrix concentrations.
Regarding the key issue of transient loading, extensive simulations have been performed of a variety of situations in which ITER divertor W monoblock (MB) melting may occur under thermal plasma impact. This is encouraged by several years of favourable experiment-theory benchmarking in current tokamaks and has been made possible by the development of a new melt dynamics code, MEMENTO, retaining much of the physics model contained within the MEMOS-U code, previously used for ITER W melt estimates. Three issues specific to the divertor have been examined: the consequences of repetitive MB top surface melting driven by uncontrolled Type I ELMs [8], the threshold for ELM-driven toroidal gap edge melting and required ELM mitigation levels in the first SRO DD H-mode plasmas [9], and the melt damage to be expected on the upper outer divertor baffle region during unmitigated downward going disruption current quenches.

References

[1] R. A. Pitts et al., Nucl. Mater. Energy 20 (2019) 100696
[2] T. Hirai et al., “Manufacturing of the ITER tungsten divertor – prototyping/qualification and status of series production”, this conference
[3] A. Loarte et al., Plasma Phys. Control. Fusion 67 (2025) 065023
[4] R. A. Pitts et al., Nucl. Mater. Energy 42 (2025) 101854
[5] J. Van Blarcum et al., to be submitted to Nucl. Fusion
[6] A. Pshenov et al., accepted for PET 2025
[7] A. Pshenov et al., Nucl. Mater. Energy 42 (2025) 101851
[8] K. Paschalidis et al., Nucl. Fusion 64 (2024) 126022
[9] K. Paschalidis et al., Nucl. Mater. Energy 44 (2025) 101955

Speaker: Richard Pitts (ITER Organization)

11:15 → 11:45 Coffee Break

11:45 → 12:25 Divertors for Next-Generation Devices

Convener: Anthony Leonard (USA)

11:45 STEP divertor design – finding a balance between multiple constraints

Aiming at building a prototype reactor based on the spherical tokamak (ST) concept by 2040, the Spherical Tokamak for Energy Production (STEP) programme has just entered the second tranche (2025-2032) which focuses on detailed design work. As part of the broader machine design effort, significant progress has been made on the divertor, addressing both physics and engineering challenges.
Based on previous work, the current design has a double-null configuration with an inner X-divertor and a tightly baffled, extended outer divertor leg. The power entering the SOL, assumed to be Psep~150 MW, is dissipated using argon (Ar) seeding both from the outer and inner legs. Due to space limitations, only the outer divertor leg is actively pumped. A dome is included to aid transfer of neutrals from the inner leg to the outer leg.
To further improve the design, we have primarily worked on the following three areas: 1) Impurity distribution (minimising the impurity fraction upstream to improve core-edge integration), 2) impact of drifts, and 3) risks during transients.
For core-divertor compatibility, we first assessed the impact of fuelling location on the Ar distribution using the SOLPS-ITER code. This demonstrated the potential to reduce the upstream Ar content by fuelling from the inner-midplane as it affects the friction/thermal forces on the Ar ions from the background plasma. Regarding He, a preliminary result of an integrated core-edge simulation predicted an intolerable helium (He) concentration in the core (~11%) without increasing the pump speed, at least by a factor 4. Using SOLPS-ITER, different pump locations were explored. This showed that pumping from the private flux region together with a vertical target geometry may reduce the upstream He concentration, albeit at the cost of more difficult detachment access (target temperature < 5eV across the entire target).
Significant effort has gone into increasing fidelity beyond existing simulations by activating particle drifts. Preliminary results with fully activated drifts for the complete set of species (D, Ar and He) show some opposite trends to earlier drift-activated simulations with evolving D and a fixed fraction of Ar, suggesting that drift effects may play an important role in Ar transport.
Transient phases, not just the flat-top/burning phase, also pose challenges. One concern is detachment front movement due to power oscillation. This has been studied with the DLS(detachment location sensitivity)-Extended, which indicated that fully detached outer legs should withstand regular transitions to SN but further design optimisation is required to improve the stability and detachment window of the inner divertor.
To rapidly evaluate STEP plasma exhaust conditions during the current ramp phases, we developed the reduced model DART (Detachment Analysis with Reduced modelling Tools), which has been validated against a large database of STEP SOLPS-ITER simulations and experiments on AUG, JET, and MAST-U. Predictions from DART indicate that high levels of Ar seeding (2-3%) are required throughout the majority of the entire plasma scenario and that the densification phase at the end the current ramp-up is a key area of concern for core-edge compatibility.

Speaker: Ryoko Osawa (UKAEA)

12:05 First-Principles–Based Divertor Optimization: A Unified MCF Divertor Framework Applied to Wendelstein 7-X

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 in stellarators and other magnetic confinement fusion (MCF) devices.

At Wendelstein 7-X (W7-X), we employ a six- σ design methodology 1— a data-driven framework that optimizes processes by quantifying a priori performance metrics within six standard deviations (σ) of process yield — combined with the Kano model [2]. Following the principle of form follows function, we categorized divertor requirements into mandatory survival criteria (e.g., resistance to heat, sputtering, and mechanical stresses) and functional performance metrics (particle exhaust and retention). These performance metrics were further decomposed into eight a priori first principles. Statistical metrics derived for each principle enable quantitative assessment of the W7-X island divertor’s current performance, shown in the table below, and facilitate direct comparisons with existing and future divertor concepts.

A field-aligned, simple SOL density model is utilized, in which perpendicular transport processes are described by a single stochastic process with a uniform perpendicular diffusion coefficient. Based on the resulting normal distribution across common and private flux region, we present seven distinct target geometries applicable to any MCF device with diverted field lines. These designs employ distinct neutral-management strategies – prioritizing attached exhaust through the SOL or PFR, or re-ionization on the incident field line, the separatrix, or SOL density peak to drive volumetric ionization losses potentially leading to higher volume recombination ratios.

A rapid modelling cycle based on anisotropic SOL diffusion EMC3-Lite modelling [3], coupled with COMSOL [4], solving the neutral transport in the molecular flow regime via the angular coefficient method and the continuous flow regime via differential equations, was established to evaluate strike line positioning and quantify the 1st, 3rd, 4th, and 5th of the 8 a priori metrics. We benchmarked these metrics for W7-X’s current divertor geometry with the 5/5, 5/4, and 5/6 resonant magnetic island configurations, and outline ongoing efforts in the W7-X divertor concept development, including the design and assessment of new divertor geometries.

This principle-driven framework bridges stellarator-tokamak divides, offering unified divertor criteria for current and next-step MCF devices. By balancing reactor demands for particle control, retention and component longevity, it advances the path toward feasible FPPs.

1 YANG, Kai; BASEM, S.; EL-HAIK, Basem. Design for six sigma. New York: McGraw-Hill, 2003.
[2] TONTINI, Gerson. Integrating the Kano model and QFD for designing new products. Total Quality Management, 2007, 18. Jg., Nr. 6, S. 599-612.
[3] FENG, Y., et al. Review of magnetic islands from the divertor perspective and a simplified heat transport model for the island divertor. Plasma Physics and Controlled Fusion, 2022, 64. Jg., Nr. 12, S. 125012.
[4] MULTIPHYSICS, COMSOL. Introduction to comsol multiphysics®. COMSOL Multiphysics, Burlington, MA, accessed Feb, 1998, 9. Jg., Nr. 2018, S. 32.

Speaker: Dr Thierry Kremeyer (Max Planck Institut für Plasmaphysik)

12:25 → 14:05 Lunch

14:05 → 14:45 Divertors for Next-Generation Devices

Convener: Anthony Leonard (USA)

14:05 Revised Divertor Design for the new European 'DEMO LAR' reactor.

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 vertical targets is maintained at 2° and the peak heat flux has been reduction during reattachment events nearly halved compared to the G1 baseline design option. This reduction ensures better manageability of the divertor under both nominal and off-normal conditions. The proposed 2D profiles should be supported by comprehensive plasma-wall interaction studies.
For these two solutions, the 3D models have been developed for a set of neutronic, thermohydraulic, electromagnetic and structural analyses for design optimization.

Speaker: Dr Giuseppe Mazzone (ENEA Nuclear Department, via E. Fermi 45, 00044 Frascati, Italy)

14:25 Power and Particle Exhaust for the Infinity Two Fusion Pilot Plant

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 confinement, elevated plasma density and high magnetic fields (B = 9 T). Two divertor designs are analyzed for Infinity Two - classical divertor and a novel large Island backside divertor (LIBD) concept which promises improved neutral pumping [2].

For the analysis performed, a suite of codes have been used; DIV3D [3] which is a diffusive field line tracing code, 3D plasma boundary code EMC3-EIRENE [4] with FLARE [5] code used to provide the input magnetic meshes, and EMC3-Lite [6] code which is the simplified version of EMC3-EIRENE that restricts the heat transport equation to only parallel electron conduction and perpendicular heat diffusion.

Results from both the divertor designs show a way to mitigate the large heat loads expected in FPP by operating with divertor detachment wherein a large fraction of power entering the scrape-off layer is radiated away. The classical divertor design possesses the flexibility in that the position of the plates can be adjusted in accordance with connection length and other parameter dependency on the target heat flux width to obtain the desired heat-flux profile. However, the open divertor geometry design leads to a poor neutral pumping efficiency. Alternatively, the LIBD results show a good particle-exhaust efficiency as the divertor design closes off pathways for neutralized plasma particles to escape back into the plasma and forces them into the divertor pumping gap. A sensitivity study for field error and beta effects has been carried out and possible ways to mitigate it has been explored.

References
[1] Hegna, C. C., et al. "The infinity two fusion pilot plant baseline plasma physics design." Journal of Plasma Physics (2025): 1-44.
[2] Bader, A., et al. "Power and Particle Exhaust for the Infinity Two Fusion Pilot Plant." Journal of Plasma Physics: 1-30.
[3] Lore, Jeremy D., et al. "Design and analysis of divertor scraper elements for the W7-X stellarator." IEEE Transactions on Plasma Science 42.3 (2014): 539-544
[4] Feng, Y, et al. "A 3D Monte Carlo code for plasma transport in island divertors." Journal of nuclear materials 241 (1997): 930-934.
[5] Frerichs, H. "FLARE: field line analysis and reconstruction for 3D boundary plasma modeling." Nuclear Fusion 64.10 (2024): 106034.
[6] Feng, Y. "Review of magnetic islands from the divertor perspective and a simplified heat transport model for the island divertor." Plasma Physics and Controlled Fusion 64.12 (2022): 125012

Speaker: Priyanjana Sinha (Type One Energy Group, TN 37931, USA)

14:45 → 15:15 Coffee Break

15:15 → 16:40 Divertors for Next-Generation Devices

Convener: Anthony Leonard (USA)

15:15 Next-Generation Modeling Approaches for Exhaust Control and Divertor Optimization

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 routine engineering tasks like real-time discharge control, scenario development, or divertor optimization. To overcome these limitations and accelerate fusion energy development, there is an urgent need for engineering-oriented modeling tools. In this work, we present two such models: a machine learning-based surrogate for real-time divertor detachment control, and a novel boundary transport model for rapid divertor optimization.

The first model, DivControlNN [1], is a data-driven surrogate designed for fast and reliable prediction of boundary and divertor plasma states using real-time control parameters and diagnostics. Trained on over 70,000 two-dimensional UEDGE simulations from KSTAR tokamak equilibria, DivControlNN employs latent space mapping [2] to efficiently capture the complex physics of divertor plasmas. This approach delivers a computational speed-up exceeding eight orders of magnitude over conventional simulations, while maintaining a relative error below 20% for key plasma parameters. As a result, it enables quasi-real-time predictions, reducing simulation times from hours to less than 0.2 ms per prediction within the Plasma Control System. A prototype detachment control system powered by DivControlNN was successfully deployed during the 2024 KSTAR experimental campaign [3]. It demonstrated robust detachment control on its first attempt, even with a new tungsten divertor configuration and without parameter fine-tuning, highlighting the model’s generalizability and immediate impact in experimental settings. The model’s predictions can be tailored to specific applications. For instance, DivControlNN forecasts plasma conditions at both the outboard midplane and divertor targets, as well as radiation information, providing key metrics for particle and heat exhaust and core plasma performance. These capabilities enable advanced integrated control systems and between-shot scenario development.

The second model aims to advance boundary transport codes suited for engineering applications by incorporating state-of-the-art numerical methods. By combining a flux-coordinate independent approach with immersed boundary techniques, this model can simulate the entire boundary plasma, from inside the separatrix to the first wall, providing critical heat and particle fluence data across all surfaces, independent of magnetic configurations. This flexibility enhances its utility for detailed divertor and blanket design, as well as integrated plasma-material modeling. Additionally, by leveraging differentiable programming, the model enables efficient sensitivity analysis and uncertainty quantification, which are essential for robust scenario development and divertor geometry optimization.

In summary, integrating machine learning surrogates like DivControlNN with new transport models marks a transformative step toward meeting the dual requirements of speed and accuracy in exhaust control and divertor design. These tools promise to accelerate the development of optimized, reliable solutions for plasma-facing components in future fusion reactors.

[1] B. Zhu, M. Zhao, et al., Physics of Plasmas, 32, 062508 (2025)
[2] B. Zhu, M. Zhao, et al., Journal of Plasma Physics, 88, 895880504 (2022)
[3] A. Gupta, D. Eldon, et al., Plasma Physics and Controlled Fusion (under review, 2025)

This work was performed for USDOE by LLNL under DE-AC52-07NA27344. LLNL-ABS-2007686

Speaker: Ben Zhu (Lawrence Livermore National Laboratory)

15:35 Bayesian Experimental Design for Divertor Heat Loads in Fusion Power Plants

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 control and off-normal-event mitigation. However, there is still no widely-used, systematic framework for the exploration, design, placement and optimisation of diagnostics to measure heat flux and temperature in fusion reactors. In this paper, Bayesian experimental design (BED) is proposed as a principled way to maximise expected information gain, thereby minimising posterior uncertainty in reconstructed quantities of interest. BED can be used to explore where to place sensors, optimise the number of sensors, and seek redundancy in diagnostic designs. We show that BED can help advise on thermocouple placement in Mega Amp Spherical Tokamak Upgrade plasma-facing components and, more generally, for thermocouple surface placement for a power plant-like divertor. A generalised methodology is offered to guide future research that employs BED with new plasma loads and divertor components.

Speaker: Michael Battye (University of York)

15:55 Discussion

16:40 → 18:00 Poster Session

16:40 Assessing the tightly baffled long-legged divertor (TBLLD) concept in TCV

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 modification to the magnetic configuration [1,2].
Simulations using the SOLPS-ITER code indicate that a TBLLD can improve TCV’s power exhaust capability by an order of magnitude compared to the unbaffled configuration [2]. Tight baffling sustains a high poloidal neutral density gradient, thereby, increasing the neutral density in front of the divertor target and enhancing volumetric power dissipation. In addition to a lower detachment threshold, extended leg length and tight baffling provide a large detachment window and a mechanism for passive detachment front stability, respectively, promising a robust power exhaust solution. The neutral cushion, futhermore, provides a reservoir of potential energy that can temporarily buffer transient loads.
The simulations informed the design of a proof-of-principle TBLLD for the outer TCV divertor. A straight, vertical divertor design enables diagnostic access via TCV’s reciprocating divertor probe array (RDPA), while maintaining engineering simplicity. Compatibility with neutral beam heated, high-power plasma scenarios constrains the baffled leg length to 0.34m. A trade-off between predicted plasma plugging and excessive recycling at the outer baffle yield a divertor width of 0.11m. The ability to expand the polodial flux along the divertor leg provides a means to vary both. The required gas tightness limits diagnostics access. Foreseen are poloidally distributed Langmuir probes, thermocouples, pressure gauges, and spectrometric lines of sight, to provide measurements of target fluxes, neutral density distribution, position and dynamics of the detachment front, which are critical to assess the TBLLD concept.
The main concerns for the proof-of-principle are the open inner divertor, which may limit the benefits of the closed outer divertor leg, and potential self-baffling of dense divertor plasmas. Recent SOLPS-ITER simulations also identified thermo-electric currents resulting from vastly different conditions in the inner and outer divertors as a potential project limitation. These concerns will be addressed through modelling and, ultimately, through experiments.
A dedicated experimental campaign with the proof-of-principle TBLLD is planned for 2026. Following a successful validation of the TBLLD concept, a second phase of upgrades will optimise the baffle geometry, extend the exhaust solution to the inner divertor, address particle exhaust, e.g. through pump ducts at the top of the TBLLD, similar to the mid-leg pumping proposed in [3], and integrate the plasma exhaust solution with an attractive core plasma scenario.
[1] M.V. Umansky, et al., Phys. Plasmas 24 (2017) 056112.
[2] G. Sun, et al., Nucl. Fusion 63 (2023) 096011.
[3] J. Yu, et al., Nucl. Mater. Energy 41 (2024) 101826.

Speaker: Holger Reimerdes (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC))

16:40 Buffering transient heat pulses with long-legged divertors in MAST Upgrade

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 space with detached divertors [3], reduced impact on the confined plasma and improved controllability of the detachment front [4]. However, data from detached conventional divertor configurations suggests their ability to buffer transient heat loads is limited, motivating the community to mitigate the source of these transients. The relatively large volume available for exhaust in alternative divertor configurations offers potentially greater capacity to buffer transient heat loads.
This contribution summarises recent experiments performed on MAST-U to study the buffering of transient heat loads arising from ELMs and sawteeth with long-legged divertor configurations within tightly baffled divertor chambers. New diagnostics have been developed to characterise the propagation and buffering of transients through the lower divertor of MAST-U in unprecedented detail, including a 500 kHz multi-channel ultra-fast spectroscopy system, divertor Langmuir probes, divertor Thomson scattering and fast visible and infra-red imaging.
The divertor ne and Te profiles prior to an ELM indicate deeply detached conditions, with Te < 1eV, ne < 1018 m-3 or both. Measurements taken as ELMs are propagating through this background plasma show an order of magnitude increase in ne and an increase in Te by factors of several up to 1ms after the ELM event. Analysis of ultra-fast spectroscopy data of the D2 Fulcher band emission across the lower divertor shows that the ratio of the Fulcher band emission from the Target and Upstream locations is well correlated with the transient heat flux arriving at the divertor plate. It also provides a valuable metric for characterising the degree to which transients are buffered by the neutral gas in the divertor chamber, where D2 Fulcher Target:Upstream emission ratios less than 1 are buffered or otherwise burn through the detachment front to the target. Measurements from a database of MAST-U pulses will be presented, containing deuterium and nitrogen gas seeding scans. Measurements performed in experiments with no impurity seeding agree well with a simplified 0D model describing the radiative dissipation of the transient heat flux via plasma-neutral interactions.
References
[1] R. A. Pitts et al., Nucl. Mater. Energy, 20 100696 (2019) [2] P. Valanju et al., Phys. Plasmas 16, 056110 (2009), [3] K. Verhaegh et al Nucl. Fusion 63 016014 (2023), [4] C. Cowley et al., Nucl. Fusion 62 086046 (2022)

Speaker: James Harrison (United Kingdom Atomic Energy Authority)

16:40 Core-edge integration studies in negative triangularity in TCV

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 reactors that circumvent H-mode challenges, including ELMs and power thresholds. In this contribution, we investigate power exhaust in Ohmic and high-power NT configurations to demonstrate the compatibility of NT plasmas with reactor relevant operation and core-edge integration.

In Ohmic L-Mode TCV discharges, detachment in NT configurations is challenging [3,4], with the outer target difficult to cool to electron temperature below 5 eV using core density ramps, compared to PT. Increasing the divertor closure with divertor gas baffles [5] decreases the outer target temperature, but detachment remains more difficult to achieve than in PT [6]. In Lower Single-Null (LSN), changing the upper triangularity (u) from positive to negative, whilst matching the divertor geometries, still results in harder detachment. This is, at least partially, explained by the role of the SOL width ($\lambda_q$), which is smaller in NT than in L-Mode PT [7,8], in agreement with theoretical and numerical predictions. These experiments with matched divertor geometries furthermore exhibit a typically lower divertor neutral pressure in NT than in PT, even with increased divertor closure [6]. Using extrinsic impurity seeding (N2), NT detachment was achieved, with reduced core confinement, and still exhibiting higher difficulty to detach as compared to PT.

These Ohmic studies have been extended to high-input power scenarios. The scenario is a 170kA LSN with favourable ion grad-B drift, with negative $\delta_u$ and positive $\delta_l$, to maintain compatibility with divertor baffles, employing Neutral Beam Heating (NBH). High performance is achieved, with $\beta_N$ up to 1.8 ($H_{98}$ near 1), at a Greenwald fraction of about 0.4, sustained in stationary conditions, for the duration of the NBH. Even without extrinsic impurity seeding, the divertor is relatively cold, as evidenced by the CIII front retreating from the outer target, a low outer target temperature (Te ~ 6 eV, measured by Langmuir probes) and a significant radiated power fraction. When $N_2$ seeding in the divertor is introduced, an X-Point Radiator forms. This, however, leads to reduced $\beta_N$. Real-time $\beta$-control can recover performance by increasing the NBH power, and enabled the demonstration of fully detached, high-performance ($\beta_N$ = 1.6) L-mode NT plasmas, comparable to ELMy H-mode, which suffers from reattachment during ELMs.

[1] CODA S., et al, Plasma Phys. Control. Fusion 64 (2022) 014004
[2] THOME K., et al, Plasma Phys. Control. Fusion 66 (2024) 206029
[3] FEVRIER O., et al, Plasma Phys. Control. Fusion 66 (2024) 065005
[4] TONELLO E., et al, Plasma Phys. Control. Fusion 66 (2024) 065006
[5] REIMERDES H., et al, Nucl. Fusion 61 (2021) 024002
[6] DURR-LEGOUPIL-NICOUD G., et al, in preparation (2025)
[7] MORGAN R., et al, submitted (2025)
[8] FAITSCH M., et al, Plasma Phys. Control. Fusion 60 (2018) 045010

Speaker: Olivier Février (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland)

16:40 Design and qualification activity of the first divertor of the DIVERTOR TOKAMAK TEST FACILITY

In the first phase of exploitation of the Divertor Tokamak Test facility (DTT) different magnetic configurations and scenarios will be study with the aim to identify the most promising. The first divertor will therefore have to be robust and flexible, able to withstand high thermal loads for long pulses and to accommodate strike points located at various positions according to the different equilibria. For this reason, almost the entire divertor plasma-facing surface is in Tungsten (W) monoblocks and the monoblocks are joined to CuCrZr pipes (plasma-facing units, PFUs) using the design and technology developed for the ITER divertor targets which, at the state of the art, is the most reliable. Furthermore, building on the design and technology developed for ITER allowed to take advantage of previous experience, limiting the necessary research and development acitvities.
The design activities started from the definition of the interfaces and of the overall constraints, which are the same ones that future divertors (the position of the rails for the fixation to the VV, the maximum size to allow insertion into the machine through the equatorial port, the maximum weight compatible with the remote handing system,....)..The poloidal profile of the first divertor followed the choice of the 3 reference magnetic equilibria (Single Null, X-Divertor and Negative Triangolarity) with which it must be compatible and the geometric and technological constraints related to the design and the production process.
The divertor consists of 3 Plasma Facing Components: Inner Target (IT), Outer Target (OT) and Central Target (CT). In each PFC there are straight parts (targets) and curved parts (baffles). In the IT and CT components there are one baffle and one target, named, respectively, Inner Vertical Target (IVT) and Central Horizontal Target (CHT), whilst in the OT there are one baffle and two targets; the Outer Vertical Target (OVT) and Outer Horizontal Target (OHT).
For each divertor module IT and CT consist of 7 PFUs, while OT of 9. The 7 central PFU of the three components are connected in series by curved tubes hidden from direct plasma radiation, but not covered with monoblocks to leave free slots between the target allowing adequate pumping close to the strike point.
With the purpose to increase the flexibility in operational scenarios by maximizing the allowable thermal load, the monoblocks have a reduced plasma side thickness to respect of the ITER ones. The monoblock thickness of 3 mm was fixed compatible both with the erosion estimates in the DTT divertor area and with the manufacturing constraints. In the development of the toroidal profile, to avoid local damage of protruding leading edges due to the gaps between PFU and to assembly and manufacturing tolerances, a toroidal monoblock bevel has been implemented. The resulting admissible load on the monoblock has been evaluated and compared with those predicted by SOLEDGE2D-EIRENE code simulation: even partially detached scenarios can be sustained by the targets in steady state for the three reference magnetic configurations. This ensures an adequate safety margin for divertor operations even in case of loss of detachment.
Although, based mainly on the well proven ITER technology, the implemented specific design solutions requested high heat flux (HHF) experimental campaigns to be validate: both HHF thermal fatigue and critical heat flux (CHF) test campaigns were conducted.
Fatigue tests [2] have proven that the monoblock with reduced thickness can withstand 1000 cycles at 20 MW/m2 without any plastic deformation and without the onset of cracks up to toroidal widths of 25 mm (maximum width for the vertical targets). Subsequently, a mock-up representative of the central target with 30 mm toroidal width was subjected to 1000 cycles at 16 MW/m² (which is the design load for this target) in the HHF test GLADIS facility at IPP in Garching. Results comparable to those of the previous campaign were found: no plastic deformation and no cracks on the surface.
The Critical Heat Flux tests were conducted in the HADES e-beam facility in the CEA center in Cadarache. The purpose of these tests was evaluating the possibility of not equipping the CT with a twisted tape in favor of design and manufacturing simplification. Considering the short plasma footprint on the CT in the NT scenarios, it was possible that the actual CHF was greater than the estimated by the semi-empirical relations. The tests were conducted on four medium scale mock-ups and two small scale ones at different water temperatures. The dependence of the CHF from the loaded area was observed, however at 60°C water temperature (reference hydraulic condition for the DTT divertor) despite the reduced loaded length, the CHF increase was marginal, highlighting the need to maintain the turbulator to ensure an adequate safety margin.
During the CHF testing (30 mm of monoblock toroidal width), deep crack formation along the tube axis was observed following complete surface recrystallization. While operational loads remain significantly below CHF levels (with a minimum safety margin of 1.4), the gradual decrease of recrystallization temperature with prolonged high-temperature exposure [3] could cause the formation of deep cracks even at lower loads. This suggested further testing to verify the survival capability of the component with a monoblock of only 3 mm of thickness even when deep cracks are presented. A mockup representative of the horizontal targets was deliberately recrystallized and loaded until cracks formed in HADES. After that, the mockup was subjected to 1000 cycles at 16 MW/m². No significant crack propagation or surface temperature variations were observed during this fatigue testing.
Finally, a small portion of the central horizontal target of the DTT divertor must be covered with flat W tiles instead of monoblocks. These flat-tiles are joined by solid state diffusion process to an OFE-Cu interlayer and subsequently joined to the CuCrZr cooling tube together with the monoblocks with the Hot Radial Pressing process. The process parameters were varied to allow the simultaneous joining of flat-tiles and monoblocks. To verify the resistance to thermal load of the developed flat-tile solution, small ad hoc mock-ups were produced and tested under thermal fatigue in the GLADIS facility.
In conclusion, the qualification activities of these years have allowed to validate most of the design choices, allowing us to reach the full-scale prototyping phase for the PFCs and the technical specifications for the cassette body and monoblocks procurement ready for the tender launching. The 1:1 scale prototype of the fixation system to the VV has already been built and is being tested and verified for compatibility with the RH system.

Speaker: Selanna Roccella (ENEA)

16:40 Development and Testing of a High-Density Tungsten Alloy Liquid Metal Divertor for Fusion Devices

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 having, 95% Tungsten and the remaining 5% are Ni & Fe has been fabricated by using powder metallurgy technique. The divertor geometry consists of seven parts, which are joined together to form a single module. The central four parts are designed in a way to form a cavity for the storage and supply of liquid lithium during operation. The texturing on the surface of alloy for the formation of wicking channels has been performed by using wire electrode discharge machining (EDM) to form a pourus capillaries to enhance the flow of liquid lithium for heat removal. The SEM and XRD of the machined samples are performed to study the roughness, cracks and phase changes at the surface after EDM machining. The same design has already been used and tested by “Magnum-PSI” for Titanium zirconium molybdenum alloy. Tantalum heater has been used for heating i.e. rising the temperature of the divertor. The heat removal efficiency of the flowing liquid metal i.e. lithium through the wicking channels has been studied.

Speaker: Asad Yaqoob Mian (Pakistan Tokamak Plasma Research Institute, PAEC Pakistan)

16:40 Divertor Turbulence Control via Leg Geometry: Experimental Tests on TCV tokamak

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 spreading is the S parameter in the Eich’s fit. While previous studies have investigated S under various conditions [1, 2, 3, 4], its behavior remains poorly constrained. A new generation of 3D plasma turbulence codes has been developed to tackle edge/SOL turbulence in diverted geometry (e.g. [5,6]), even revealing unexpected results under reactor-relevant conditions [7], highlighting the challenges of extrapolating current tokamak experiments to reactor scenarios in the edge and SOL regions.
In this contribution we present an exprimental test for such codes, based on the theoretical ideas proposed in [8]. The key idea is that the relative orientation of the magnetic curvature vector κ and the pressure gradient ∇p determines the turbulence transport into the private flux region of the two divertor legs, depending on their geometry. This alignment can either destabilize or stabilize drift-interchange and drift-wave turbulence. Controlling the turbulence drive and its impact on transport through diverted plasma geometry represents a promising approach in the design of alternative divertor configurations. Leveraging the shaping flexibility of the TCV tokamak, a scan of the outer divertor leg orientation has been performed while maintaining matched upstream plasma parameters and target flux expansion. Initial results from Langmuir probes and infrared analysis do not show any significant difference in the divertor spreading parameter S, while fast camera analysis is still ongoing. However, according to [8], the inner divertor leg is expected to be more affected, as the κ and ∇p drifts are parallel when the leg is vertically oriented. A database analysis is currently underway, based on discharges in negative triangularity, to further investigate this effect. Finally, taking advantage of the new upper divertor shaping capabilities at ASDEX Upgrade, similar but less extensive scans are planned for the ASDEX Upgrade tokamak as well.

[1] M. A. Makowski et al, Phys. Plasmas 19, 056122 (2012)
[2] B. Sieglin et al 2013 Plasma Phys. Control. Fusion 55 124039
[3] M. Roberto, Phd Thesis, EPFL (2020)
[4] D. Brida et al 2025 Nucl. Fusion 65 026065
[5] A. Stegmeir et al 2018 Plasma Phys. Control. Fusion 60 035005
[6] S Ku et al 2006 J. Phys.: Conf. Ser. 46 87
[7] C.S. Chang et al 2017 Nucl. Fusion 57 116023
[8] N. Walkden, et al. Commun Phys 5, 139 (2022)

Speaker: Marco Cavedon (Department of Physics, University of Milano-Bicocca, Italy)

16:40 DIVGAS: A Robust and Reliable Workflow for Modelling Divertor Neutral Gas Flow and Optimizing Divertors in Next-Generation Fusion Reactors – Applications to DTT and EU-DEMO Divertors

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 users to choose the best fit for their needs. The stochastic module is based on the Direct Simulation Monte Carlo (DSMC) method, while the deterministic module on the Discrete Velocity Method (DVM). Both methods accurately capture neutral gas behaviour across the entire range of collisionality expected in fusion reactor particle exhaust systems, establishing DIVGAS as the leading tool for divertor research. DIVGAS is focused on one of the most critical challenges in fusion reactor operations: creating an approach for particle exhaust in the divertor that aligns with both operational constraints and requirements. DIVGAS represents the main numerical tool for particle exhaust simulations within several EUROfusion work packages (i.e. WP-TFV, WP-PWIE, WP-DIV, etc.).
In the present work two representative divertor analysis examples will be presented, namely the DTT (Divertor Tokamak Test Facility) [1-3] and the EU-DEMO (European Demonstration Power Plant). Both examples include 3D complex modelling of sub-divertor neutral gas dynamics, marking them as the first of their kind, and emphasize the significance of three-dimensional effects. Several key quantities are quantified, including particle fluxes through various surfaces, while the contours of neutral pressure and temperature are also provided to offer a comprehensive overview of the neutral gas dynamics in the sub-divertor region. For both divertors, a significant impact of the leakages on pumping performance was observed, with deuterium flux through the leakages contributing approximately 20-30%. Meanwhile, the flux through the entry gaps towards the plasma vessel consistently accounts for the largest portion of the total incoming flux to the sub-divertor area (~60% for DTT and >60% for EU-DEMO). The present DIVGAS simulations uncover key insights into the particle exhaust of DTT and DEMO, providing a robust framework for developing optimization strategies to improve divertor pumping performance.

References
[1] C. Tantos et al., Nucl. Fusion 62 026038 (2022); [2] C. Tantos et al., Nucl. Fusion 64 016019 (2024); [3] C. Tantos et al., Fusion Eng. Des. 115021 (2025)

Speaker: Dr Christos Tantos (Karlsruhe Institute of Technology (ITEP))

16:40 Effects of active divertor cryo-pumping on particle exhaust in Wendelstein 7-X

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 their areas of intersection with the divertor targets, leading to strike lines on different parts of the divertor and thus different distributions of neutral gas pressure in the subdivertor volume. In standard configuration, effective particle confinement times of 5-10\,s were determined regardless of the state of the cryo-vacuum pumps. Due to higher neutral gas pressures at otherwise similar plasma parameters, lower effective particle confinement times could be reached in the high iota configuration, leading to better particle exhaust. Additionally, a significant effect of cryo-vacuum pumping on the effective particle confinement times could be determined with the mean effective particle confinement time decreasing by more than half from 8.7\,s during discharges without cryo-vacuum pumping to 4.0\,s during discharges with cryo-vacuum pumping. Similarly, the recycling coefficient in standard configuration was determined to be on average 0.97 and remains unaffected by cryo-vacuum pumping, whereas the mean recycling coefficient of 0.98 reached in high iota configuration without cryo-vacuum pumping is reduced to 0.95 when the cryo-vacuum pumps are used. \
%Additionally, density control during discharges using only cryo-vacuum pumping has been demonstrated for long %plasma discharges with steady-state pellet injection.\
DSMC (Direct Simulation Monte Carlo) simulations of the neutral gas pressure in the subdivertor volume using the DIVGAS computational platform confirm the decrease of the neutral gas pressure by up to 11$\%$ when using the cryo-vacuum pumps and thus confirm the improved particle exhaust realized by cryo-vacuum pumping in the high iota configuration [2].

[1] O Grulke et al. “Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled
plasma facing components”. In: Nucl. Fusion 64.11 (2024), p. 112002.
[2] S Varoutis et al. “Numerical analysis of gas exhaust in Wendelstein 7-X using the Direct Simulation
Monte Carlo method”. In: Nuclear Fusion (2025).

Speaker: Victoria Haak (Max-Planck-Institut für Plasmaphysik)

16:40 Impurity seeding for the control of the radiated power using bolometers in Wendelstein 7-X stellarator device

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. Volumetric power dissipation by radiation in the plasma edge reduces the peak heat flux on the PFCs. However, regulating the radiated power requires delicate control. Low radiated power compromises the integrity of the divertor, while excessively high radiated power can provoke an earlier performance degradation. For this reason, a feedback control system is necessary. This contribution discusses the recently operational feedback control system for the radiated power in the Wendelstein 7-X stellarator device (W7-X).
In W7-X the radiation feedback control system was a crucial upgrade. It allowed reliable machine operation despite the challenges with the target/baffle loads in high performance and power conditions. The impurities are seeded by piezo-electric valves [3]. The radiated power proxy is calculated with a specialized wide-angle bolometer camera of only 5 channels [4]. Using feed-forward impurity injection experiments with multi-sinusoidal amplitude modulation allows us to characterize the radiation frequency-resolved response to the actuator [5]. The contribution shows the parameter dependency for density, radiated power fraction (f_rad) and heating power of the transfer functions of the system. The transfer functions are simplified expressions, which provide the timescales and gains, with which the system reacts. The controller design methodology demonstrates the applicability and improvements of the approach for optimized radiation feedback control, which helps to avoid oscillatory behavior.
The controllers operated successfully, allowing for reliable control of the radiated power proxy, both with Nitrogen and Neon. The optimized controller achieves good accuracy with an overshoot in a controller step on the order of 10% and a rise time of less than 200 ms. The feedback system in W7-X allowed reliable operation and accurate control across wide operation parameters. This included challenging high power scenarios and stable detachment (f_rad≈85%)) in a range of W7-X magnetic configurations. The feedback system enabled efficient implementation of long-pulse operation (320 sec,1.2 GJ) in detached conditions and successfully corrected for radiation drifts, likely due to changes in wall outgassing. The remaining challenges for reliable control in the 3D stellarator are toroidal asymmetries in the radiation response due to seeding that have been newly discovered.
[1] N. L. Cardozo, Fusion on the back of an envelope, Lecture Notes TU/e 2022
[2] A. W. Leonard, Plasma Phys. Control. Fusion 60 044001 (2018)
[3] M. Griener et.al, Rev. Sci. Instrum. 88, 033509 (2017), https://doi.org/10.1063/1.4978629
[4] G. Partesotti et.al, Rev. Sci. Instrum. 95, 103503 (2024); doi: 10.1063/5.02
[5] R. Pintelon et al, System Identification: A Frequency Domain Approach (IEEE Press, 2001)

Speaker: Mr Anastasios Tsikouras (Max Planck Institute for Plasma Physics, Greifswald, Germany)

16:40 MODELLING DIVERTOR SOLUTIONS FOR POWER EXHAUST: IN-DEPTH EXPERIMENTAL VALIDATION IN TCV

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 both approaches, allowing for flexible divertor shaping [2] and variable gas-baffling [3] while featuring extensive divertor diagnostics.

This work presents a comprehensive validation of SOLPS-ITER [4,5] simulations against a broad experimental database built from L-mode discharges in TCV, including standard lower single null (LSN) and advanced divertor configurations. The experiments were specifically designed for code validation, relying on high reproducibility and maximised edge and divertor diagnostic coverage. This enables two-dimensional measurements of electron and ion temperatures, density, parallel ion flow, and impurity emissivity, along with wall heat and particle loads and divertor neutral pressure.
Legacy SOLPS-ITER simulations were found to overestimate dissipative effects at the targets, predicting a denser and cooler divertor than observed experimentally [6]. This work introduces key modelling refinements: core-edge coupling is improved using the JINTRAC framework; ion flux limiters are investigated to account for kinetic effects; and carbon source modelling is revised through improved treatment of physical and chemical sputtering. These adjustments lead to significantly improved agreement with experiments, reproducing target conditions and divertor neutral pressure across various density regimes, levels of divertor closure, and magnetic geometries.

Insights from validated scenarios have informed predictive modelling of TCV’s forthcoming divertor upgrade, the Tightly Baffled Long-Leg Divertor (TBLLD). Preliminary simulations indicate enhanced detachment and impurity control [7], supporting its development as a promising advanced divertor concept for future fusion devices.

---

[1] B. P. Duval et al Nucl. Fusion 64 (2024) 112023
[2] C. Theiler et al Nucl. Fusion 57 (2017) 072008
[3] H. Reimerdes et al 2021 Nucl. Fusion 61024002
[4] S. Wiesen et al J. Nucl. Mater. 463 (2015) 480
[5] X. Bonnin et al Plasma and Fusion Research 11 (2016) 1403102
[6] M. Wensing et al Phys. Plasmas 28 (2021) 082508
[7] G. Sun et al Nucl. Fusion 63 (2023) 096011

Speaker: Elena Tonello (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC))

16:40 Network identification of the design structure matrix with application to STEP

Design structure matrices (DSMs) represent the connections between elements composing a system. In fusion, they are used to visualize the dependencies between various plasma variables, processes, states, and events and they serve as a basis for synthesizing supervisory controllers. A DSM representing the existence and absence of these plasma relations has been constructed after strong consultation of experts [1]. The disadvantage of building the DSM purely on the basis of expert knowledge is that it only captures exactly that what these experts think is important.

In this work, we show that a DSM can also be obtained from data only, by applying dynamic network identification methods. The plasma is viewed as a network in which the dependencies between the plasma properties are represented as the edges between nodes. The interconnection structure of the network is equivalent to the DSM and can be identified from sensor and actuator data, without explicitly estimating the exact dynamics. We extend the Bayesian model selection method of [2,3] to include excitation signals and use this data-driven network method to identify the interconnection structure.

We apply this method to a simulation example, where we identify the particle flow within the five-chamber tokamak model of STEP. The plasma in the reactor is modelled as a dynamic network with five inputs (the gas flux inlets) and 12 measured outputs (the number of neutral and plasma particles). The internal relations between the measured variables are identified with the network identification method. As a result, the DSM representing the gas and plasma particle flow within the tokamak is identified from simulation data. This illustrates the potential of using dynamic network theory for identifying DSMs from data.

References
[1] T.F. Beernaert, ''Discovering the potential of dependency structure modelling for fusion systems engineering'', PhD thesis, Eindhoven University of Technology, 2024.
[2] S. Shi, G. Bottegal, and P.M.J. Van den Hof, ''Bayesian topology identification of linear dynamic networks'', In Proceedings of the 18th European Control Conference (ECC), pages 2814–2819, 2019.
[3] SYSDYNET, ''Matlab app and toolbox for dynamic network identification'', Control Systems Group, Eindhoven University of Technology, 2024, www.sysdynet.net.

Speaker: Lizan Kivits (DIFFER)

16:40 On the Possibility of Monitoring Liquid Metal Plasma-Facing Components via Emission by Computer Vision

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 this, the direct observation of LM PFC with optical emission spectroscopic (OES) diagnostics towards the testing site has been found difficult. Fortunately, it is still possible to have at least a visible digital camera to monitor the LM PFC in some devices, e.g. EAST. In EAST [6-7], it has been discovered that the line emission of lithium (Li) was strong, emphasizing the intensive interaction of Li species with plasma electrons. Even though the direct OES diagnostics is none, the reconstruction of line emission by image processing, together with the survey OES diagnostics can show the emission trends of LM vapor interacting with electrons. The initial sample line spectra of lithium (Li) and tin (Sn) can be plotted using information from literature, e.g. the relative line intensities by NIST’s Handbook of Basic Atomic Spectroscopic Data [8]. This sample line spectra can be decomposed to a few important eigenbases with respect to sample line emission wavelength [9], keeping the significant emission trends. The amplitude factor for each eigenbasis can be estimated by the RGB information of digital images after performing the transformation to the CIE1931 xyz color space. With this, the trends of the reconstructing line emission are compared to the survey OES diagnostics. This method allows qualitatively investigating LM behavior at edge plasmas and conceptualizes the possible monitoring tools for LM PFCs used in future fusion devices.

References
[1] Nygren R and Tabarés F 2016 Nuclear Materials and Energy 9 6–21
[2] Saenz F et al 2022 Nuclear Fusion 62 086008
[3] Ruzic D et al 2017 Nuclear Materials and Energy 12 1324–1329
[4] Vertkov A V et al 2019 Inorganic Materials: Applied Research 10 326–332
[5] Zuo G et al 2014 Fusion Engineering and Design 89 2845–2852
[6] Zuo G et al 2017 Nuclear Fusion 57 046017
[7] Zuo G et al 2022 Nuclear Materials and Energy 33 101263
[8] NIST 2024 Handbook of Basic Atomic Spectroscopic Data NIST Standard Referece Database 108
[9] Otsu H et al 2018 Computer Graphics Forum 37(6), 370‒381

Speaker: Nopparit Somboonkittichai (Department of Physics, Faculty of Science, Kasetsart University, Thailand)

16:40 Predictions of radiation front location, volume, and radiated power dependency on operational parameters in a mid-leg pumped “chimney” divertor

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 ($\rm T_{e,Xpt}$ ~ $\rm T_{e,sep,OMP}$) simultaneously with a detached divertor target (Te,targ < ~5 eV). UEDGE simulations, including magnetic and $\rm\bf E\times B$ drifts, predict radiation front control between the pumping plenum and X-point using low-field side main-ion gas puffing with the ion $\rm \mathbf{B}\times\nabla B$ directed towards the X-point. In the opposite B-field direction, with ion $\rm \mathbf{B}\times\nabla B$ directed away from the X-point, limited control is predicted due to $\rm\bf E_\theta\times B$ and $\rm\mathbf{E}_r\times \bf B$ divertor drift flows yielding significant current to the pumping surface, resulting in efficient removal of the injected gas.

This work will expand on previous simulations of a proposed mid-leg pumped low-field side “chimney” divertor, planned to be installed in the DIII-D upper divertor [1,2]. The “chimney” design employs a closed divertor structure and an extended outer leg with a 4.5 cm deep volume at the end of the divertor slot, which acts as a neutral reservoir for power exhaust through ion-neutral interaction [3]. The UEDGE-predicted sensitivity of the radiation front location, volume, and power to operational parameters, including main- and impurity particle injection and applied heating power will be presented. Furthermore, UEDGE model fidelity will be increased compared to initial simulations [1] by modeling the transport of individual charge states of the seeded impurities and including fluid molecules as a separate species in the simulations. The UEDGE pumping model applied will be compared to DEGAS2 kinetic Monte-Carlo neutral simulations to assess the fidelity of the UEDGE fluid neutral pumping model and validate previous results [1].

The predictive UEDGE and SOLPS-ITER simulations used to design the divertor will be validated against experimental measurements using the proposed chimney divertor and associated diagnostic systems in DIII-D. If passive radiation front stabilization and location control consistent with edge-plasma code predictions are observed experimentally, mid-leg pumped divertors may present an alternative divertor design for future divertors with reduced requirements on divertor volume and radiating impurity injection.

[1] A. Holm et al., “Modeling a divertor with mid-leg pumping for high-power H-mode scenarios in DIII-D considering E × B drift flows,” Nuclear Materials and Energy, vol. 41, p. 101782, 2024.
[2] J. H. Yu et al., “Simulations of divertor designs that spatially separate power and particle exhaust using mid-leg divertor particle pumping,” Nuclear Materials and Energy, vol. 41, p. 101826, 2024.
[3] R. S. Wilcox, et al., “The “Chimney” divertor: A closed divertor with mid-leg pumping for core-edge integration in DIII-D”, this meeting.
This work was supported in part by the US Department of Energy under contract nos. DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725. LLNL-ABS-2006456.

Disclaimer: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Speaker: Andreas Holm (Lawrence Livermore National Laboratory)

16:40 Progress regarding tungsten-copper metal matrix composites based on additive manufacturing for plasma-facing components

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 so-called monoblock design, which uses W armor blocks joined to a Cu alloy heat sink pipe. This design has demonstrated good damage resilience, but is not readily scalable because of the high number of individual W monoblocks needed for a reactor-scale divertor PFC. Another known approach is the so-called flat-tile design, which comprises flat W armor tiles bonded to a bulk Cu alloy heat sink. While the flat-tile design exhibits a higher heat removal capability and follows a more straightforward manufacturing route than the monoblock design, the thermal expansion mismatch between monolithic W and Cu causes high thermomechanical stresses that can lead to delamination of armor tiles from the heat sink. Against this background, the contribution presents an advanced concept for PFCs, utilizing additive manufacturing (AM) processes to create W-Cu metal matrix composites (W$_{AM}$-Cu MMCs). Exploiting the design freedom of AM, W$_{AM}$-Cu MMCs can be tailored to minimize thermally induced stresses in a PFC under HHF loading. Results regarding the fabrication and HHF testing of small-scale PFC mock-ups based on such an approach will be presented. The investigated mock-ups are based on AM W preforms that were built by means of a powder-bed fusion process and subsequently infiltrated with Cu utilizing a vacuum-assisted melt infiltration process. These mock-ups were then tested in the GLADIS HHF test facility at the Max Planck Institute for Plasma Physics (IPP) in Garching. The contribution presents results of HHF tests on four different W$_{AM}$-Cu MMC mock-up specimens including accompanying FEM simulations and post-exposure metallographic investigations.

Speaker: Robert Lürbke (Max-Planck-Institut für Plasmaphysik)

16:40 Simulation study of the effect of impurities on the ELM dynamics using BOUT++ six-field two-fluid model

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 to understand the mechanism how impurities affect the dynamic of ELMs. The “indirect” effects of impurities, such as radiative cooling, profile regulation and fuel dilution, have been investigated in the simulation studies by the linear instability analysis [3-5]. However, it is still insufficient to explain the complicated ELM behaviors observed in experiments [6, 7].
In this work, a systematical simulation study of the dynamical effect of impurities on pedestal stability, ELM evolution and turbulence transport is carried out using BOUT++ six-field two-fluid module [8] combined with the impurity model developed by Li et al. [9]. According to Seto et al.’s work [10, 11], the evolutions of toroidal axisymmetric electric field and parallel current are included. The electron inertia is also considered for self-consistent current dissipation in fast magnetic reconnection.
Based on EAST equilibrium (shot #91616, t = 6s during ELM suppression phase), the nonlinear simulations are performed. It is found that the complete suppression of ELM can be achieved only when including the effect on vorticity distribution due to impurities, which reflects the essential role of impurity dynamics in ELM suppression. Further scan of the impurity density shows the change of ELM size with impurity density and indicates that there could an impurity density “threshold” for ELM suppression, which is qualitatively consistent with experiment result in HL-2A [7].
The impurity effect on the ELM dynamic process is further explored by varying both the impurity mass ratio and the pedestal electron temperature. As the simulated ELM evolution shows the “two-stage burst” [11] feature, it is found that the first stage is related to linear trigger of peeling-ballooning mode and the second stage is nonlinear triggered due to the drift-tearing mode [12]. According to the difference in the “two-stage burst” feature, the two-dimensional parametric plane can be divided into four regions. According to the shift of operation point between different regions, the different effect of impurity on the ELM evolution can be understood.

Reference
[1] A. W. Leonard. Plasma Physics and Controlled Fusion 60 (2018) 044001.
[2] G. S. Xu, et al. Reviews of Modern Plasma Physics 7 (2023) 14.
[3] H. Lan, et al. Nuclear Fusion 60 (2020) 056013.
[4] T. H. Osborne, et al. Nuclear Fusion 55 (2015) 063018.
[5] J. Rapp, et al. Plasma Physics and Controlled Fusion 44 (2002) 639.
[6] X. Lin, et al. Physics Letters A 431 (2022) 127988.
[7] G. L. Xiao, et al. Nuclear Fusion 61 (2021) 116011.
[8] T. Y. Xia, et al. Nuclear Fusion 53 (2013) 073009.
[9] Z. Li, et al. Nuclear Fusion 62 (2022) 076033.
[10] H. Seto, Physics of Plasmas 26 (2019) 052507.
[11] H. Seto, Physics of Plasmas 31 (2024) 032513.
[12] A. B. Hassam, Physics of Fluids 23 (1980) 2493.

Speaker: Shifeng MAO (University of Science and Technology of China)

16:40 SOLPS-ITER Modeling of Low Recycling Divertor Solutions for the Spherical Tokamak Advanced Reactor (STAR)

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 those that would be experienced during the transient event. So long as the emission from the PFC can be limited [3], or the plasma flow can be controlled such that there is sufficient screening from the main plasma [4], liquid metals could minimally affect the upstream plasma. Recently, a variant of the lithium vapor box divertor concept [3,5] has been proposed for NSTX-U, only requiring a small evaporating region in the private flux region (PFR) to control the target heat flux. Furthermore, a Capillary Porous System with Fast flowing liquid lithium (CPSF) [7] has been proposed as a potential PFC, with SOLPS-ITER modeling indicating steady state peak target heat fluxes of 18 MW/m2 could be handled with PFC peak temperature ~715°C in NSTX-U [3]. Modeling of these concepts are hindered by the lack of certainty in the deuterium recycling rate, known to be high recycling at higher temperatures (~400°C), though with a strong dependence on the vessel pressure [8]. To demonstrate feasibility of these liquid lithium PFC concepts in a reactor and explore the effect of deuterium recycling reductions on the scrape-off layer, SOLPS-ITER is employed to model the Spherical Tokamak Advanced Reactor (STAR) [9]. STAR is a pilot plant concept, designed to take advantage of improved confinement from spherical tokamaks, while leveraging liquid metals to control the power exhaust. SOLPS-ITER results for STAR indicate that acceptable steady-state PFC temperatures can be achieved while handling PSOL > 135MW. The upstream impurity concentration resulting from employing these designs on STAR is found to vary significantly with the assumed divertor recycling coefficient. A temperature-dependent, radially varying model for the recycling is also introduced, which indicates a transition to a high-recycling regime above 350°C-400°C, in agreement with experimental data [8]. Combination with an external radiator is also tested and shown to have beneficial effects at controlling the upstream concentration of lithium, via reduced heat flux on the lithium PFC. Upstream lithium concentration ~0.02 is shown to be possible while maintaining Zeff ~ 2.4.
[1] F. Maviglia et al. Nuclear Materials and Energy 26 (2021) 100897
[2] P. Rindt et al. Nuclear Fusion (2019) 59 054001
[3] E.D. Emdee et al. Nuclear Fusion (2024) 64 086047
[4] E.D. Emdee and R.J. Goldston Nuclear Fusion (2023) 63 096003
[5] R J Goldston et al. 2016 Phys. Scr. (2016) 014017
[6] E.D. Emdee and R.J. Goldston Nuclear Materials and Energy 41 (2024) 101737
[7] A. Khodak and R. Maingi Physics of Plasmas 29 (2022) 072505
[8] M. Morbey et al. Nuclear Fusion 64 (2024) 076019
[9] T.G. Brown and J.E. Menard Fusion Engineering and Design 192 (2023) 113583

Speaker: Eric Emdee (Princeton Plasma Physics Laboratory)

16:40 Study on the Thermal Performance of ITER Tungsten Divertor Monoblock Using Nanofluid for Cooling Enhancement

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 divertor which can sustain very high heat fluxes. In a previous work, the author developed a mathematical model to investigate the steady state and transient thermal–hydraulic performance of ITER tungsten divertor monoblock. The model could predict the thermal response of the divertor structural materials for bare cooling tube and a cooling tube with swirl-tape insertion. Nanofluids have gained extensive attention due to their role in improving the efficiency of thermal systems. Azmi et al. report a further enhancement in heat transfer coefficients in combination with structural modifications of flow systems namely, the addition of tape inserts. In this work a mathematical model has been developed/updated to investigate the thermal performance of the ITER tungsten divertor monoblock using new heat transfer enhancement technique. In order to enhance the heat transfer process, a water based TiO2 nanofluid at 3% concentrations is used to cool the divertor. The model is then used to predict the steady state thermal behaviour of the divertor under incident surface heat fluxes ranges from 2 to 20 MW/m2 for a nanofuild cooled tube with swirl-tape insertion as well as water cooled bare and swirl-tape tubes. The operating conditions are: inlet temperature: 150°C, pressure: 5 MPa and coolant velocity: 16 m/s. Calculations are performed for incident surface heat flux of 2, 4, 6, 8, 10, 12, 14, 16, 18 and 20 MW/m2. Fig. 1 shows the variation of the predicted maximum tube-surface temperature values versus the incident heat flux for a divertor of bare tube, swirl-tape tube and swirl-tape tube cooled by nanofluid. It shows that, for bare tube divertor, themaximum tube wall surface temperature exceeds the ONB temperature for incident heat fluxes greater than 10 MW/m2and so subcooled boiling is predicted at the top surface of the tube, and for swirl-tape tube divertor, subcooled boiling is predicted at the top surface of the tube for incident heat fluxes greater than 18 MW/m2. On the other hand, the combined effect of swirl-tape insertion and nanofluid shows a maximum tube-surface temperature lower than the ONB temperature by a considerable margin even at an incident surface heat flux of 20 MW/m2.

Fig. 1. Maximum tube-surface temperature.

Speaker: Salah El-Din El-Morshedy (Prof. Dr. of Thermal-hydraulics, Egyptian Atomic Energy Authority)

Wednesday 29 October

09:00 → 10:45 X-point and other Radiator Regimes

Convener: Felix Reimold (Max Planck Institute for Plasma Physics, Greifswald, Germany)

09:00 Topical Session 2 - X-point and other radiator regimes

Speaker: Felix Reimold (Max Planck Institute for Plasma Physics, Greifswald, Germany)

09:05 Overview of XPR experiments and gaps in extrapolating to future devices

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 MAST-U \citep{Soukhanovskii_2022}, but only at low or moderate heating powers. An outstanding feature of ASDEX Upgrade (AUG) is its high heating power ($\ge$\ 20 MW) compared to its size ($R=1.65$\ m). In order to study a variety of ADCs \citep{Lunt_2017} at these high power conditions AUG has installed a pair of in-vessel divertor coils, a charcoal coated cryo-pump capable of capturing He, new divertor targets as well as an outstanding set of diagnostics in its upper divertor. During the two-year long opening enormous technical challenges were solved, like the in-vessel winding of the continuous conductor or the installation of the divertor tiles with an alignment accuracy of 0.2 mm. Since April 2025 the new advanced upper divertor is now fully operational and the experimental campaign in full-swing. In a first step several ADCs like the X-divertor (XD, \citep{Lunt_2019b}), the Low-Field-Side Snowflake minus (LFS SF$^-$, \citep{Pan_2018,Pan_2020}) or an extreme form of the Compact Radiative Divertor \citep{Lunt_2023} have been established. As an example Fig.\ \ref{fig:LFSSFm} shows an infrared thermography image recorded during the LFS SF$^-$ phase of a high power discharge. The overplotted magnetic equilibrium shows the primary (red) and secondary (purple) separatrices. The example clearly shows the SOL power splitting between the primary and secondary outer strike lines. In a second step, power, fueling and impurity seeding scans were performed to study the exhaust and detachment behavior of the different configurations. We present the first results of these experiments with heating powers of up to 20 MW, interpret them by means of SOLPS and/or EMC3-EIRENE modelling and give an outlook to further studies.

Speaker: Tilmann Lunt (Max-Planck-Institut für Plasmaphysik)

09:45 On the effect of mixed impurity seeding on occurrence and stability of the X-point radiator in fusion reactors

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 different tokamaks including JET [1]. XPR stability and the fraction of powerdissipated by radiation can be actively controlled through impurity seeding and neutral gas fuelling [2]. The control parameter is the XPR height, which represents its vertical extension above the X-point. Controlling the XPR height is also important to prevent marfes and consecutive disruptions.

This contribution focuses on the prospects of XPR scenarios in tokamak fusion reactors and the appropriate choice of impurity species. The investigations indicate that XPRs should be more accessible in a reactor than in smaller tokamaks, i.e., at lower neutral and impurity densities. This is consistent with results from SOLPS-ITER transport code simulations [4]. Among the impurities investigated (B, C, Ne,
N, Ar), argon is most efficient at radiating height power. For EU-DEMO, argon seeding resulted in an XPR height of 50 cm, which corresponds to a dissipation of 90 % of the 150 MW heating power. However, the XPR height depends strongly on impurity concentration, which could make control difficult. Neon, on the other hand, requires higher concentrations to radiate the same amount of power, but the XPR hight changes more slowly with concentration. Therefore, an impurity mixture
could be advantageous, with argon carrying the base load of radiation losses and neon being used for control. Under the model assumptions made, there are indications that carbon and, to a certain extent, nitrogen have a negative effect on XPR stability, while too much boron may lead to unstable XPRs.

These results were obtained from a reduced power and particle balance model [5] and an extension of it [6] which estimates the XPR height, the dissipated power, and the coupling to the upstream profiles. The calculated reduction in the pedestal gradient is consistent with the experiments [8] and could explain the process of ELM
suppression.

[1] M. Bernert et al., Nuclear Materials and Energy 43, 101916 (2025).
[2] M. Bernert et al., Nucl. Fusion 61, 24001 (2020).
[3] T. Lunt et al., Phys. Rev. Lett. 130, 145102 (2023).
[4] O. Pan et al., Proc. of the 50 EPS Plasma Physics Conference, Spain, 2024.
[5] U. Stroth et al., Nucl. Fusion 62, 076008 (2022).
[6] U. Stroth et al., Plasma Phys. Contr. Fusion 67, 025001 (2025).
[7] O. Pan et al., Nucl. Fusion 63, 016001 (2022).
[8] E. Wolfrum et al., Proc. of the 50 EPS Plasma Physics Conference, Spain, 2024.

Speaker: ULRICH STROTH (MPI für Plasmaphysik)

10:05 Progress towards JINTRAC integrated modelling of X-point radiators

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 known that ELMs must be mitigated or even entirely suppressed to ensure sufficient target lifetime and W density control [4]. To date, modelling work regarding the XPR regime has been focused on divertor and scrape-off layer characteristics with, in particular, several demonstrations of stable XPR solutions obtained with the SOLPS-ITER plasma boundary code suite for present devices and ITER [5,6]. However, for burning plasmas, a critical question is the core-edge compatibility of these XPR regimes. Such assessments require integrated modelling and efforts in this direction being pursued at the ITER Organization (IO) are the subject of this contribution.
Our approach is to deploy and, where necessary, further develop the modelling code suite JINTRAC, currently the workhorse for high-fidelity integrated simulations at the IO (used extensively, for example, in the recent ITER re-baselining physics activities [7]). As a first step, we focus on existing published SOLPS-ITER stationary XPR simulations obtained for ASDEX Upgrade [5] and try to reproduce them with standalone runs of the JINTRAC divertor/SOL model EDGE2D-Eirene. By utilizing identical grids, a benchmark has been performed in the first instance without drifts and currents or neutral-neutral collisions and with Dirichlet conditions applied at the core boundary on plasma temperatures and ion density. We demonstrate that EDGE2D-Eirene is able to obtain quantitatively similar solutions for the XPR in these nitrogen seeded cases.
As a second step towards the integrated model, we also show that a time-dependent XPR can be obtained using Neumann boundary conditions on the core boundary, where the heat and particle flux from the core are prescribed. Herein, we find that at sufficiently high $X_A$ (i.e. the XPR access parameter [8]), the temperature just above the X-point rapidly drops to a few eV. This indicates bifurcation behaviour between a ‘hot’ and ‘cold’ X-point solution as described in the work of Stroth [8]. The impurity radiation steadily rises through the transition, but sees no rapid transition as the temperature above the X-point. Without feedback on nitrogen seeding, the XPR tends to develop over time into an uncontrollable MARFE. A controller is being implemented to actively control the height of the XPR to obtain a fully stationary simulation result, in the same way that the XPR is stabilised in experiment [1]. Regarding the ultimate aim of the full integrated model for the XPR regime, we will discuss the ongoing efforts required to make this possible within the JINTRAC framework. In particular, the need for the 2D boundary code to also describe the pedestal region whilst normally this is taken care of by the core codes in JINTRAC.

[1] M. Bernert et al., Nucl. Fusion 61 (2021) 024001
[2] T.O.S.J. Bosman et al., Nucl. Fusion 65 (2025) 016057
[3] V.A. Soukhanovskii et al., Nucl. Mat. Energy 41 (2024) 101790
[4] R.A. Pitts et al., Nucl. Mat. Energy 42 (2025) 101854
[5] O. Pan et al., Nucl. Fusion 63 (2023) 016001
[6] A. Poletaeva et al., Nucl. Fusion 64 (2024) 126038
[7] A. Loarte et al., PPCF 2025 accepted manuscript, DOI:10.1088/1361-6587/add9c9
[8] U. Stroth et al., Nucl. Fusion 62 (2022) 076008

Speaker: Sven Korving (ITER Organization)

10:25 The X-Point radiating regime at JET in D and DT plasmas with mixed impurities

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 almost all currently operating tokamaks and was, in JET, first observed in 2015 [1]. In the recent JET campaigns, which culminated in the final DT campaign (DTE3) and the subsequent shutdown of the machine, the XPR, along with its stable control, was successfully demonstrated and investigated in detail.
Several seed impurities were injected in order to trigger an XPR, such as nitrogen, neon, argon, and combinations thereof. With pure neon or argon seeding the plasma exhibits dithering between H- and L-mode, even at heating powers of up to 26 MW. The dithering is suppressed and the plasma stays in H-mode when combining two impurities or with pure N2 seeding. As the use of N2 was not permitted in the DT campaign, the mixture of Ar and Ne performed best while still being compatible with DT operation. SOLPS-ITER simulations were conducted using N2, Ne, Ar, and a Ne+Ar mixture. XPRs were achieved with all impurity options; however, the Ne case, compared to N2 and Ar, exhibited less impurity compression in the XPR region and a broader distribution of the radiative mantle, as well as a less bifurcation-like transition when entering the XPR regime. The XPR access conditions, radiative capabilities, stability, and impact on upstream parameters for different impurities were compared and analyzed using a combination of theoretical models [3,4] and SOLPS-ITER simulations.
For the first time at JET, a movement of the XPR was tracked by the horizontal bolometer camera and provided in real time to the control system, using the algorithm developed at AUG [2]. A PI controller is implemented using Ar seeding as actuator (while the Ne injection is pre-programmed). The reaction time of the XPR location to a change in the seeding rate is in the range of 1 s, much slower than at AUG, which presents a restriction for the XPR control. However, external perturbations, such as drops in heating power or pellet injection, are first buffered by a movement of the XPR, and can then be effectively counteracted by the slower control. The active control of the XPR helped to efficiently establish the same power exhaust conditions when moving from D to DT plasmas. The overall performance of the scenario, though initially low without seeding (H98≈0.65), increases slightly when going to DT and does not decrease with impurity injection. Notably, the edge kinetic profiles are not observed to be affected by the strong seeding, while ELMs become fully mitigated.
[1] M. Wischmeier, et al., J. Nucl. Mater. 463, 22-29 (2015).
[2] M. Bernert, et al., Nucl. Fusion 61, 024001 (2020).
[3] U. Stroth, et al., Nucl. Fusion 62, 076008 (2022).
[4] D. Morozov and A. Pshenov, Plasma Physics Reports 41, 599 (2015).

Speaker: Ou Pan (Max-Planck-Institut für Plasmaphysik)

10:45 → 11:15 Coffee Break

11:15 → 12:15 X-point and other Radiator Regimes

Convener: Felix Reimold (Max Planck Institute for Plasma Physics, Greifswald, Germany)

11:15 Impact of plasma scenario on W erosion and migration: comparison between WEST experiments and SOLEDGE3X-ERO2.0 simulations

A major issue for next step devices is the control of plasma wall interaction, both for keeping the material erosion at reasonable level as well as for avoiding core contamination by high Z impurities and consequently the reduction of plasma performances. In this respect, WEST experiments supported by numerical modeling are particularly relevant to progress in the physical understanding of the complex interplay between erosion patterns, impurity migration and efficiency of plasma screening. In this contribution we will review main results from the comparison between WEST experimental data and SOLEDGE-ERO2.0 numerical modeling, both in 2D and 3D.
First, we will present results concerning the high fluence campaign with attached plasma conditions that was conducted in WEST to expose the ITER-grade actively-cooled divertor to ITER-relevant deuterium fluences. The same plasma discharge of 60s long was repeated hundreds of times, accumulating about 10,000 seconds of plasma with a maximum of 6×10^(26) part.m−2 of fluence measured by divertor Langmuir probes (LP). Impurities have been tracked by visible spectroscopy (VS), showing high content of nitrogen and boron all along the campaign, with oxygen and carbon also present. The estimation of W gross and net erosion is estimated both considering analysis LP and VS experimental data and simulation results, and finally compared with post mortem analysis. Numerical simulations are particularly helpful in determining the concentration of light impurities necessary for explaining the measured net erosion as well as the impact of ExB drifts on the asymmetries between inner and outer strike point with a quite important level of deposits at the inner strike point.
In order to reduce W sources and core contamination that have been found in attached plasma, X-Point Radiator (XPR) and detached regimes are particularly attractive. In WEST, XPR have been obtained with nitrogen seeding for discharges of more than 30 seconds long. Very low level of W erosion at the strike points is measured during these discharges using visible spectroscopy, as previously predicted with SOLEDGE-ERO2.0 simulations for equivalent target conditions. Moreover, thanks to the movable VUV spectrometer, we have recently measured strong reduction of W low ionized states around the X-point during XPR with respect to attached conditions. These measurements provide important information on the migration of the eroded W and are particularly valuable to validate the predictions of SOLEDGE-ERO2.0, encouraging the development of such scenario for improving control strategies on W erosion and core contamination for future experiments.
Finally, we will present the results of SOLEDGE-ERO2.0 simulations considering both 3D wall and magnetic geometries. In particular the impact of magnetic ripple on the modulation of the plasma outflow on the divertor target has been compared with experimental data. Its impact on W erosion and migration estimated with ERO2.0 is ongoing and it will be also presented.

Speaker: Guido CIRAOLO (CEA, IRFM)

11:35 Experimental characterization and numerical investigation of X-point radiator regime in WEST

The X-point radiator (XPR) regime, observed in the WEST tokamak, is characterized by the appearance of a stable radiative ring above the X-point with sufficient nitrogen seeding. The transitions from high-recycling regime to the XPR regime usually happen within very short time. XPR regime can significantly mitigating the heat load at the divertor target while maintaining core plasma performance, which offers a promising approach for advanced tokamak operation and control. This study employs the SOLEDGE3X-EIRENE code in transport mode to numerically reproduce the XPR radiation regime under WEST boundary conditions. Simulations were conducted with a constant input power up to 2 MW, deuterium gas puffing combined with nitrogen seeding, and included full drift effects. The simulation results shows that nitrogen seeding, introduced either from the midplane or the private flux region, can lead to the formation of an XPR inside the separatrix, like in experiments. When XPR occurs, one observes a significant reduction in the target temperature, dropping from above 10 eV to 2 eV, while the upstream conditions remain nearly unchanged. The density profile, initially asymmetric with higher values at the inner target and lower values at the outer target, becomes symmetric. As the seeding rate increases, the XPR shifts toward the low-field side along the magnetic field line, consistent with experimental observations. Further analysis indicates that drifts may play a critical role in the formation and dynamics of the XPR. During the onset of the XPR, a potential well develops at the X-point. Changes in the potential map and the leading reversed E×B flux are strongly correlated with the abrupt system transitions observed during the cliff-like behavior associated with XPR onset. The dynamics of the cliff-like transition from an attached plasma to the XPR state is also correctly reproduced and analysed with the help of reduced bifurcation models.

Speaker: Eric SERRE (CNRS)

11:55 X-POINT TARGET RADIATOR AS A POTENTIAL REACTOR EXHAUST SOLUTION

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 plasma core, effectively preventing radiative edge cooling. This spatial separation promises attractive core–edge integration, avoiding proximity to operational limits and possible degradation of core performance. A comprehensive diagnostic suite, including high-resolution spectral imaging, bolometry, Langmuir probes, and Thomson scattering, reveals the two-dimensional structure of the XPTR. In Ohmic plasmas, compared to standard single-null (SN) configurations, detachment access is significantly facilitated, with divertor target heat fluxes reduced by more than a factor of five. This is accompanied by the formation of a stable, localized radiative zone, exhibiting a marked insensitivity of the radiator position to upstream density changes, increased by a factor >5 compared to the SN, facilitating detachment control.
Following the initial demonstration of the XPTR in [1], its excellent exhaust performance has also been confirmed in scenarios with high levels of second-harmonic ECRH, with values of the detachment accessibility metric $PB/R/n_{sep}^2$ comparable to those predicted for SPARC. While nitrogen seeding in this scenario does not lead to detachment in SN, it triggers early XPTR onset and associated detachment in the X-point target divertor. These studies, initially focused on the lower divertor, are now being extended to up–down symmetric (double-null) geometries and are supported by reduced analytical models and first SOLPS-ITER simulations with full drift effects included.
Overall, the XPTR offers several compelling advantages for reactor applications, particularly supporting the planned integration of the X-point target divertor in the SPARC reactor [3] and the ARC pilot plant. These advantages include the realization of a robust detached divertor regime with reduced reliance on impurity seeding, simplified control requirements, and minimal risk of confinement degradation from excessive radiative edge cooling.
[1] K. LEE, C. THEILER, M. CARPITA et al., Phys. Rev. Lett. 134, 185102 (2025)
[2] B. LABOMBARD et al., Nucl. Fusion 55, 053020 (2015)
[3] A. Q. KUANG et al., J. Plasma Phys. 86, 865860505 (2020)

Speaker: Prof. Christian Theiler (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), Lausanne, Switzerland)

12:15 → 13:50 Lunch

13:50 → 15:35 X-point and other Radiator Regimes

Convener: Felix Reimold (Max Planck Institute for Plasma Physics, Greifswald, Germany)

13:50 First attempts to establish an X-point radiator regime on MAST-U in N-seeded H-modes

This presentation summarises the first attempts to develop an X-point radiator (XPR) regime on MAST-U. The XPR, a localised radiating structure near the magnetic X-point, is a promising power exhaust solution for future fusion reactors. While XPR regimes have been demonstrated on most operating devices, MAST-U remains one of the last major tokamaks yet to access this regime. The work presented will further aim to additionally assess the effect of open versus closed divertor geometries and the influence of magnetic topology (DN vs. LSN) on XPR formation and stability.

The experiments have so far focused on double-null plasmas with conventional divertor geometry, using N2 seeding into beam heated ELMy H-mode scenarios. Figure 1 presents time traces from a one-beam (1.5 MW) scenario which was recently developed with improved initial confinement (H₉₈ ~0.8). As the gas is injected, H₉₈ decreases slightly to ~0.75. The ELMs transition from Type I to smaller grassy Type III ELMs. As shown by the 2D poloidal image of total radiation in Figure 1, this transition coincides with inner divertor detachment and radiation localised strongly within the X-point and weakly along the outer divertor leg, suggesting incomplete outer leg detachment. As the scenario evolves, ELMs become fully suppressed at t=0.27 s; however, the loss of the density pedestal indicates a transition not to a sustained ELM-free H-mode, but to an H-L back transition. This is likely triggered by the arrival of the q = 1 surface, causing rotation locking between the q = 1 and q = 2 surfaces. Notably, the confinement only decreases significantly after the mode completely locks during the ELM phase at t=0.32.

Seeding has also been attempted in a two-beam (~3 MW) scenario, albeit this older scenario had poorer initial confinement (H₉₈ ~0.7). The radiation measurements in this scenario indicate complete detachment of both divertors, with radiation strongly emitting around the X-point and along the inner core separatrix. During the nitrogen seeding phase, ELM behaviour transitions from Type III to small grassy ELMs, while H-mode is sustained with a strong density pedestal. The H₉₈ factor decreases modestly from ~0.7 to ~0.6. Ongoing experiments aim to further investigate XPR access and stability in the high-performance scenario using two-beam heating.

Speaker: Stuart Henderson (UKAEA)

14:10 Characterization of X-point radiation operational space and performance impact in DIII-D H-mode discharges

X-point radiation experiments in the DIII-D tokamak explored stable X-point radiating (XPR) conditions with mitigation of edge localized modes (ELMs) potentially relevant to steady-state divertor operation in future devices while collecting detailed local measurements of plasma parameters and species-resolved radiating emissivities for model validation. Regimes with X-point radiation [1] are being pursued to simultaneously achieve deep detachment to maintain manageable target heat fluxes and intrinsic mitigation of ELMs. While such regimes have been achieved in many tokamaks (JET, ASDEX Upgrade, TCV, DIII-D), an improved understanding of operational access requirements, radiation stability as well as impact on overall confinement and impurity dilution are critical for the extrapolation of these regimes to future devices.
Dedicated experiments (H-mode, favorable ion B$\times \nabla$B drift direction, $I_p$=1.3 MA, $P_{INJ}$= 6-12 MW) were performed in the DIII-D tokamak for a broad characterization of access to X-point radiation regimes, validation of radiation stability models including dependence on radiating species, and assessment of the impact of X-point radiation on pedestal and confinement. Stable X-point radiation was accessed from deeply detached conditions via feedforward impurity seeding from the private flux region (deuterated methane $CD_4$, and nitrogen $N_2$). Access to XPR regimes was accompanied by a reduction in confinement up to 20% compared to deeply detached conditions, with a decrease in pedestal temperatures at nearly unchanged pedestal densities. Deeper X-point radiation resulted in a back-transition to L-mode while maintaining radiation inside the X-point, without unstable evolution into a MARFE. A narrower operating space was observed with only C as the dominant radiator, consistently with theory [2]. While theory predicts more restrictive conditions for X-point radiation due to unstable MARFE evolution, experimentally the limited operating space was identified in terms of a narrower window of access to X-point radiation before back-transition to L-mode. Radiated power densities remained concentrated in the last 5% of normalized poloidal flux $\psi_N$ in the confined plasma. Divertor Thomson scattering measurements inside the X-point indicate Te $\sim$1-2eV with up to a $5-10\times$ reduction in electron pressure with respect to upstream pressure in the confined plasma in the last 1% of $\psi_N$. Penetration of the radiation front inside the X-point was accompanied by ELM mitigation from $\Delta W_{ELM}/W \sim 1.2-1.5\%$ to $\Delta W_{ELM}/W \sim 0.3-0.5\%$ with an increase in ELM frequency from 100 Hz to $\sim$300Hz. While the power lost to the scrape-off layer (SOL) via ELMs remained on the order of 10-20% of $P_{SOL}$, the energy lost per ELM and the peak divertor heat fluxes were largely reduced by $4\times$ and $10\times$ respectively, indicating increased ELM buffering. At the largest seeding rates ($\sim$20 Torr l/s), dilution measured at the pedestal top can become large with both carbon ($f_C$) and nitrogen ($f_N$) concentrations simultaneously up to 3%. Work is ongoing to compare detailed local measurements of plasma parameters and impurity concentrations to simple models and edge fluid codes to contribute to the physics basis of X-point radiating regimes towards establishing their potential as core-edge integration solution for future devices.
[1] M. Bernert et al 2021 Nucl. Fusion 61 024001
[2] U. Stroth et al 2022 Nucl. Fusion 62 076008

This material was supported by the U.S. Department of Energy under Awards DEAC52-07NA27344, DE-FC02-04ER54698, DE-AC05-00OR22725, DE-NA0003525.

Speaker: Filippo Scotti (LLNL)

14:30 Experimental research on the deeply detached X-point radiator regime on EAST

Radiative divertor detachment with impurity seeding is considered one of the most promising means for mitigating particle and heat fluxes on the divertor target. To measure the impurity radiation distribution, a tangentially viewing camera system for lower divertor plasma observation has been developed and installed on EAST. A reconstructed 2D distribution of N II line radiation is obtained based on Phillips-Tikhonov regularization, revealing the electron temperature region in the range of 6-10 eV during a nitrogen (N2) seeding experiment.
With N2 seeding and radiation feedback control, the deep detachment on the lower outer (LO) divertor target with the stable X-point radiator (XPR) has been maintained for ~ 5 s on EAST. The profiles of electron temperature (Tet) and heat flux (qt) with distance to strike point larger than 6 cm (ρ ~ 1.06, larger than λjs) are radially flat on the outer divertor target in the deeply detached state. The peak qt on the LO divertor target in the deeply detached state decreases by more than 80%, compared with the attached state. The strong XPR contributes to the heat flux control on the divertor target and has a good stability. In addition, there are significant decreases of tungsten and boron radiation in the LO divertor region, which indicate an effective reduction of divertor sputtering. In the deeply detached state on the LO divertor target, there is a confinement degradation with decreases of electron temperature, density and pressure in the core region. The above results show that XPR is a promising detachment operation regime for divertor protection in future reactors.

Speaker: Kedong Li

14:50 Discussion

15:35 → 16:05 Coffee Break

16:05 → 17:30 Divertor Engineering and Materials

Convener: Richard Pitts (ITER Organization)

16:05 Topical session 3 - Divertor Engineering and Materials

Speaker: Richard Pitts (ITER Organization)

16:10 Tungsten and copper (alloy) based divertor target plasma-facing components – state-of-the-art and developments towards the application of tungsten-copper composites

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 properties of PFC materials, like conductivity or toughness. Divertor target PFC designs have been devised for next generation MCF devices, especially ITER and DEMO-like devices. These solutions are based on joining a monolithic tungsten (W) armour to a high-conductivity heat sink based on copper (Cu) alloy, which ensures an acceptably high heat removal capability of the PFC. However, the use of monolithic W and Cu represents a technological challenge, in particular in terms of the thermal expansion mismatch that can lead to high thermomechanical stresses, which are a persistent driving force towards the failure of a PFC under cyclic high heat flux (HHF) loading. In this context, the demanding boundary conditions for next generation and reactor-scale devices outlined above ask for the development of advanced material and design solutions for divertor PFCs with improved performance and durability. One promising route regarding this further development is the use of composites, aiming at sophisticated combinations of W and Cu products that mitigate known design concerns through targeted PFC material reinforcement and macroscopic property tailoring. Against this background, the contribution will summarise the current state-of-the-art regarding solid W and Cu (alloy) based divertor target PFC designs, as adopted for ITER and foreseen to be used in DEMO-like devices. Further, advanced design options based on W-Cu composites will be discussed, especially including concepts based on the reinforcement of PFC materials with high-strength W fibres, as well as the use of tailored composite structures realised through additive manufacturing (AM). These approaches have undergone several years of development and have been evaluated by means of HHF testing on PFC mock-ups. In addition, related activities regarding the industrial upscaling of composites fabrication will be discussed.
Keywords: plasma-facing component, tungsten, copper alloy, composite, fibre-reinforcement, additive manufacturing

Speaker: Alexander von Müller (Max-Planck-Institut für Plasmaphysik)

16:50 Comparative Overview of Divertor Cassette Design and Integration in EU-DEMO, VNS and DTT

The design of the divertor is one of the most critical challenges in the development of fusion reactors due to various intensive driver loads and dependence on physics modelling. It must exhaust a significant fraction of the plasma heating (mainly from fusion alphas) which is conducted through the Scrape-Off Layer (SOL) during normal, transient, and off-normal events, while also protecting the vacuum vessel from nuclear loads and providing a plasma-facing surface compatible with intense plasma–material interactions. Managing power exhaust in the divertor region is therefore a central challenge for reactor-scale fusion devices, requiring advanced engineering solutions, intensive technological developments and testing, and the careful integration of subcomponents under stringent thermal, structural, and operational constraints.
This contribution presents a comparative overview of the divertor design approaches adopted for three major European tokamak devices currently under development within the framework of Eurofusion activities: EU-DEMO, the Volumetric Neutron Source (VNS), and the Divertor Tokamak Test facility (DTT). While the EU-DEMO divertor is primarily driven by large-scale reactor dimensions and stringent plant availability requirements, the divertors of VNS and DTT share similar geometric scales but diverge significantly in terms of magnetic configuration flexibility and neutron load conditions. Starting from the main design drivers and boundary conditions for each machine, we examine both commonalities and specific challenges across the three concepts, with a focus on geometric integration, heat flux management, coolant operating conditions, structural and electromagnetic (EM) loads, and compatibility with remote handling (RH) systems. This comparative analysis aims to highlight synergies across the three concepts, and to explain how design choices are shaped by machine mission, performance targets and operational context.

Speaker: Domenico Marzullo

17:10 W/Cu divertor development at ASIPP towards fusion reactor through the projects of BEST and CRAFT

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 requirements in 2016 [1]. In 2021, the EAST lower divertor was also upgraded with W/Cu components [2]. Both upper and lower W/Cu divertors significantly improve the capability of high-power operation, contributed to the many achievements in plasma research such as the 1066 s H mode operation. In recent years, W/Cu divertor technology have being oriented to the development of meter-scale large components for fusion reactor, mainly including, 1) design exploration for continuous high heat flux removal in the closed V corner for BEST (Burning-plasma Experimental Superconducting Tokamak) divertor, 2) qualification of meter-scale BEST divertor components with monoblock structure and conventional materials of pure tungsten and CuCrZr, and 3) development of flat-type components with advanced materials in CRAFT.
A closed corner slot design was proved to be benefit for the detachment operation while keeping a high performance of core plasma [3]. This corner slot design prefers to have continuous heat flux removal capability around the corner slot, which introduce significant challenges in engineering design [2]. For this purpose, such a divertor design was initially explored with the W/Cu hypervapotron as plasma facing unit (PFU) [4]. However, due to the rigid connection of horizontal target and vertical target in the closed V corner, such design has very high thermal stress which introduce the risk in the life of thermal fatigue. Therefore, much efforts were paid to decouple the water-feeding structure for PFU from the mechanical connection between the horizontal and vertical stainless-steel support for plasma facing components (PFCs). A final design with W/Cu monoblock PFU was achieved with the continuous heat flux removal of 15 MW/m2 on both horizontal and vertical targets in the closed V corner. This closed V shape PFU design was further developed and integrated into the PFCs and a 2-m large divertor module for BEST tokamak. The BEST divertor module has a new design for cooling circuit that can effectively reduce the pressure drop without any negative impact on the divertor by rationally allocating the coolant through outer target to inner target and dome [5].
The continuous heat flux removal 15 MW/m2 lead to a sophisticated design of W/Cu PFCs, which increase the difficulty in the fabrication and assembly of BEST divertor. To study the engineering feasibility, at first a W/Cu PFU mock-up was made and successfully tested by 5000 cycles of 15 MW/m2 without visible failure. A lot of small R&Ds was done, such as the multi-materials tube-to-tube joint by e-beam welding. Afterward, the meter-scale long W/Cu PFU was made for inner and outer target. The full-scale prototype of BEST divertor module has been fabricated and assembled, and now is subjecting to the various testing including the coolant pressure, leak tightness after baking and high heat flux.
Another important work is the development of fusion reactor flat-type divertor with advanced materials, i.e., potassium-doped tungsten (KW) for armor, oxide dispersion-strengthened copper (ODS-Cu) alloy for heat sink, and reduced activation ferritic/martensitic (RAFM) steel for structure [6]. The small to middle scale mock-ups as well as prototypes for CFETR have been successfully tested with 1000 cycles of 20 MW/m2 in CRAFT project.

[1] G.N. Luo, et al., Overview of decade-long development of plasma-facing components at ASIPP, Nucl. Fusion 57 (2017) 065001.
[2] Y.T. Song, et al., Recent EAST experimental results and systems upgrade in support of long-pulse steady-state plasma operation, IEEE Tran. Plas. Science, VOL. 50, NO. 11, 4330-4334, 2022
[3] G.S. Xu, et al., Physics design of new lower tungsten divertor for long-pulse high-power operations in EAST, Nucl. Fusion 61 (2021) 126070.
[4] S.Q. Feng, et al., Development of continuous V-shaped structure for high heat flux components of flat-type divertor, Nuclear Materials and Energy 35 (2023) 101419
[5] X.Y. Qian, et al., New designs of target and cooling scheme for water cooled divertor in DEMO, Nucl. Fusion 61 (2021) 036008.
[6] Tiejun Xu, et al., Design and Research and Development of Front-Face Remote Handling Targets for CFETR Divertor, IEEE Tran. Plas. Science, VOL. 52, NO. 9, 3542-3548, 2024

Speaker: Mr Xuebing PENG (Institute of Plasma Physics, Chinese Academy of Sciences)

Thursday 30 October

09:00 → 10:45 Divertor Engineering and Materials

Convener: Richard Pitts (ITER Organization)

09:00 Manufacturing of the ITER tungsten divertor – prototyping/qualification and status of series production

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 European DA the IVT and the Cassette Body, with the Russian DA supplying the Dome. Following a successful prototype phase, all of the principal components are now in the series production phase.
The VT Plasma-facing Units (PFU) have adopted the W monoblock concept with swirl tape in straight parts [2] whereas those of Dome, which do not need to tolerate the highest stationary heat fluxes, use the flat tile concept on a hypervapotron heat sink [3]. The PFUs are mounted on the Steel Support Structure and welded to the coolant channels. Following several years of design consolidation, the procurement phase began by the qualification of manufacturing technologies to produce full-scale prototypes, delivering, for example, 2 m long PFUs of unprecedented size. During this process, strict requirements were imposed on qualification, manufacturing and inspection. The quality of W to heat sink joint, dimensional compliance and leak tightness, are the key requirements. Technical assessment on the requirements based on experience from the prototype manufacturing has been performed in view of the process optimization, such as tolerances, manufacturing strategy in machining, attachment and welding. Series production had been started in the qualified supplier sites.
The manufactured components are to be delivered to the integration site where the Divertor Cassette Assemblies will be in the final configuration. In parallel, integration of divertor mounted diagnostics where applicable will be performed. They are to be accepted only after a series of factory acceptance tests.
On delivery to the ITER site, the Divertor Cassette Assemblies will be installed through three lower ports into the vacuum vessel (VV) by divertor cassette movers. They will be inserted radially and then transported toroidally to the final position, where the cassettes are positioned and pre-loaded on the toroidal rails. Custom machining of the toroidal rail elements ensures the alignment requirements with respect to the Tokamak Assemblies Datum allowing for as-built VV dimensions.
This contribution will provide an overview of the ITER divertor procurement progress, technological solutions and return of experience from prototype manufacturing. The current plans Divertor Cassette integration and installation in the VV will also be described.
[1] M. Merola, et al., “Engineering challenges and development of the ITER Blanket System and Divertor” Fusion Eng. Des. 96–97, 2015, 34-41.

[2] T. Hirai, et al., “ITER divertor materials and manufacturing challenges”, Fusion Eng. Des. 125 (2017) 250–255.
[3] T. Hirai, et al., “Hypervapotron Heat Sinks in ITER Plasma-Facing Components – Process Qualifications and Production Control toward Series Production” Fusion Eng. Des. 187 (2023) 113454 (6 pp).

Speaker: Dr TAKESHI Hirai (ITER Organization)

09:20 Technology qualification of W based water-cooled target modules for Wendelstein 7-X

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 plasma-facing materials to ensure low tritium retention. Therefore, a project was launched in 2021 to develop a next generation divertor with a plasma facing surface made of W based material. One task is to optimize the geometry of the plasma-facing surface to prevent overloads, improve particle exhaust and maximize impurity retention away from the core plasma. In parallel, a second task is the structural design of a W based target module and the qualification of suitable manufacturing technologies, which is conducted in the framework of the EUROfusion funded WPDIV program. Similar design constraints are imposed as for the current CFC divertor in terms of design heat load (10 MW/m²), plasma-facing surface shape, cooling water supply and target module size and weight.
This paper presents the progress on the qualification of the manufacturing technologies of the design presented in [Fellinger: 2023]. The main technology under qualification is an additively manufactured CuCrZr heat sink made using laser powder bed fusion (LPBF) with a functionally graded plasma-facing coating of W/Cu or WNiFe/Cu, applied by cold gas spraying or low-pressure plasma spraying. Alternative to a coating, a mosaic of so-called sandwich tiles is diffusion bonded or brazed onto the heat sink. These flat tiles exist of W or WNiFe onto which a thin soft copper layer is galvanized or diffusion welded. As the normal direction of the tiles varies over the plasma-facing surface, an external uniaxial jacket cannot apply contact pressure during bonding. Therefore, the Cu side of the tiles are diffusion bonded using a paste with nano-sized Cu particles at moderate temperatures (< 550°C) and an external pressure that can be realized by springs or other technologies. Alternatively, the tiles are brazed at high temperature (~980°C) without pressure using a Cu/Au braze foil. Notably, Ni89P11 and Ag based brazes have been successfully been tested in high heat flux (HHF) tests in the past [Tokitani: 2021, Böswirth: 2024], but P is not compatible with CuCrZr and Ag is not allowed in W7X.
The paper shows results of thermal and mechanical properties of WNiFe and LPBF made CuCrZr, and the He leak tightness of LPBF samples. Furthermore, microstructural analyses are presented on bonding trials with nano-paste, coated samples and on sandwich tiles before and after HHF tests. Finally, thermal and hydraulic design simulations were performed and confirmed by results of hydraulic tests and HHF tests on LPBF made CuCrZr heat sinks. These tests demonstrated that the heat sink can handle the design heat load with robust margin towards the allowed pressure drop and CuCrZr limit temperature.

Speaker: Joris Fellinger (Max-Planck-Institute for Plasma Physics)

09:40 Simulation of DTT disruptions with JOREK: benchmarking MaxFEA results and assessing divertor electromagnetic loads

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 plasma current tokamak such as DTT are plasma disruptions, which can drive high transient heat loads and electromagnetic forces and lead to damage to in-vessel components (IVC). Generally, the workflow leading to the structural verification of any given IVC design must consider (in addition to all operational loads) the most severe plasma disruption event. Therefore, starting from the disruption simulation results and using them as input for the electromagnetic (EM) analysis, the EM loads acting on the component can be evaluated, and its structural integrity assessed.
DTT IVCs have been verified using, as input data for the EM calculations, the results of MAXFEA simulations [1]—a multi-platform finite element code. MAXFEA can model axisymmetric disruptions by solving the Grad-Shafranov equilibrium equation for the plasma together with Maxwell’s equations and Ohm’s law in passive and active conductors. However, in MAXFEA, the current density profile and its evolution are predetermined, which prevents the prediction of both the current quench (CQ) duration and the halo width, parameters that are crucial for the accurate calculation of EM loads. The work reported here is focused on benchmarking MAXFEA disruption simulations using JOREK [2] a state-of-the-art non-linear MHD code for plasma simulations, capable of predicting these parameters by solving transport equations for the thermal energy and current density, providing a physics-based benchmark for the simulations performed with MAXFEA.
The results of the MAXFEA unmitigated disruption simulations, used as input for the verification workflow of the DTT Divertor have been compared with the disruptions computed using JOREK. As a first step, all the required DTT data were compiled into the Integrated Modeling and Analysis Suite (IMAS), the software framework to be used for all physics modeling and analysis at ITER (including with the JOREK code) [3] .This also constitutes the beginning of a longer term, wider effort to integrate all future DTT physics analysis and hopefully experimental data eventually obtained from the machine into the IMAS framework. For the purposes of the study described here, DTT data integrated in IMAS include: the machine description (including the central solenoid, poloidal field coils, vacuum vessel, divertor-first wall, and all passive structures considered in a 2D plasma equilibrium/disruption) and the initial equilibrium scenario (5.5 MA plasma current and 6 T toroidal magnetic field) with all the relevant plasma core profiles. JOREK has been initialized using the IMAS data, leading to the simulation of an unmitigated, slow downward Vertical Displacement Event (VDE) with a CQ duration of tCQ = 50 ms and a mitigated VDE with a tCQ = 24 ms. The predictive JOREK results were compared with the initial assumptions taken by MAXFEA, showing very good agreement with the global quantities that drive the evaluation of the EM load.
Key words: IMAS integration, DTT, Disruptions, Divertor,
This work was carried out in the frame of a Cooperation Agreement between DTT and ITER. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization

References:
[1] P. Barabaschi, The MAXFEA code, Plasma Control Tech. Meet. Naka (1993).
[2] M. Hoelzl et al., Nucl. Fusion 61 (2021) 065001
[3] F. Imbeaux et al., Nucl. Fusion 55 (2015) 123006

Speaker: Mr Nicola Massanova (University of Trieste)

10:00 Discussion

10:45 → 11:15 Coffee Break

11:15 → 12:20 Scrape-off-Layer and Divertor Physics

Convener: Marco Wischmeier (IPP Garching)

11:15 Topical Session 4 - Scrape-off-Layer and Divertor Physics

Speaker: Marco Wischmeier (IPP Garching)

11:20 Linking power exhaust constraints with the Separatrix Operating Space through multi-machine (ITER, JET, SPARC) modelling and experiments

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 necessitate the need for connecting SepOS projections to power exhaust constraints defined by divertor target parameter specifications for tolerable PFC heat and particle loads. Quantifying the strong reduction in the separatrix density with increasing impurity concentration, as shown in the ITER divertor physics basis SOLPS projections [1], is of particular interest as it introduces another variable in the SepOS-power exhaust optimization.

In this work we focus on the interpretation of multi-device boundary plasma simulation datasets with the aim of framing dissipative divertor properties in terms of reduced dimensionality models that can be used to impose power exhaust constraints on the SepOS. We focus on exploiting observed self-similarities in SOLPS-ITER simulation datasets of ITER, JET and SPARC to inform experimental strategies in reactor-scale devices to safely map out the integrated scenario operational space. The strategy is underpinned by a strong dependence of divertor momentum and power losses on the target electron temperature, Te,t, describing a range of ‘dissipation pathways’ over a range of fueling, impurity seeding, heating power and cross-field diffusive transport assumptions. The variation in the momentum and power loss trends across the three devices constrains the achievable upstream parameters and thus the accessible dissipative SepOS. Within this framework, we quantify the reduction in upstream electron density, ne,sep, as a consequence of strong radiative losses in impurity seeded scenarios and contrast the numerical results against JET-ILW experiments which show significant (30-50%) ne,sep reductions in L-mode and H-mode high frad experiments. Lastly, using synthetic diagnostic techniques, we consider observables that satisfy reactor compatibility constraints while preserving the fidelity of the simulation output physics quantities, with emphasis on passive divertor spectroscopy.

[1] Pitts R.A. et al 2019, Physics basis for the first ITER tungsten divertor, Nuclear Materials and Energy 20 100696

Speaker: Bart Lomanowski (Oak Ridge National Laboratory)

12:00 On the validation of dynamic models for (exhaust) control on DEMO class devices; a comparison in multiple fidelities to experiments in TCV

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 necessitates a control system (not just for the exhaust) that guarantees operation within safety-critical limits in presence of disturbances. Such guarantees rely on dynamic models of the entire plasma that: 1) capture the response to actuators and disturbances; 2) connect to reactor relevant sensors; 3) describe safety critical limits; 4) scale to DEMO size reactors.

In this contribution we validate the dynamics in multiple physics-based models using a system-identification experiment in the TCV tokamak [2]. This data class has proven its use in controller design, providing guarantees for stability and performance [3]. We compare: 1) TCM a three chamber model simulating 0D reservoirs [4]; 2) DIV1D using reservoirs combined with a 1D scrape-off layer [5]; and 3) SOLEDGE3X-EIRENE as the high-fidelity 2D plasma-edge simulator [6]. We find that the coupling to a core reservoir and a realistic time-scale for neutrals to ionize allows the models to align with the measurements. Similar validations on other devices (using existing data) should be prioritized over predicting behavior on non-existing devices.

To enable model-based control for DEMO-class devices, the challenge for integrated modeling is to shift from interpretations (on isolated domains) to full device time-dependent simulations that mimic the complexity faced when operating DEMO. We should prioritize quantification of errors in solutions with respect to the operational window, disturbances and control-relevant dynamics. Physics-based models should be used to design and demonstrate controllers (that take advantage of real-time data) with operational guarantees jointly for core and exhaust. To be clear, demonstration means running high-performance detached discharges without ELMS or disruptions with the 99% reliability required in DEMO, ARC, STEP, and ITER.

References
[1] A. Loarte et al. Plasma Physics and Controlled Fusion, 2025.
[2] B. Duval et al. Nuclear Fusion, 64(11):112023, 2024.
[3] T. Ravensbergen et al. Nature Communications, 12(1):1–9, 2021.
[4] J. Koenders et al. 41st Benelux Meeting on Systems and Control, 2022.
[5] G. Derks et al. Plasma Physics and Controlled Fusion, 66(5):055004, 2024.
[6] H. Yang et al. Plasma Physics And Controlled Fusion, 65(12):125005, 2023.

Speaker: Gijs Derks (DIFFER - Dutch Institute for Fundamental Energy Research, Eindhoven, Netherlands and Eindhoven University of Technology, Control Systems Technology, Eindhoven, Netherlands)

12:20 → 14:00 Lunch

14:00 → 15:20 Scrape-off-Layer and Divertor Physics

Convener: Marco Wischmeier (IPP Garching)

14:00 Status of Divertor Operation in Stellarators

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 divertor performance metrics are the same for tokamaks and stellarators: heatfluxes to the divertor within engineering limits <10 MWm$^{-2}$, $T_{div}$<5 eV to minimize W sputtering, impurity screening, and sufficient pumping of Helium ash to avoid core dilution/sufficient pumping of fuel particles to maintain density control. Thus far, experiments on W7-X, utilizing the island divertor with open divertor geometry, show deficiencies both in heat exhaust at higher input powers (in at least one main configuration) [1] and in density build-up at the target for fuel/Helium pumping.
This contribution summarizes the latest developments and work towards advancing the physics basis of the stellarator divertor. The stellarator scrape-off layer (SOL) is significantly more complex than that of a tokamak: the magnetic connection lengths are generally much longer due to the lower field line pitch, increasing the weight of perpendicular transport [2]. Additionally, the inherent 3D shape of the flux tubes and discontinuous divertors result in complicated relationships of the plasma parameters along the field lines [3]. Both of these aspects of the 3D SOL make it impossible to use either simplified models or sophisticated modeling capabilities developed for tokamaks in their present form. Of particular importance is understanding the effects of drifts and anomalous cross-field transport processes in determining divertor performance: measurements in W7-X indicate that the poloidal flow velocities between the X-points, largely thought to be attributable to drifts, are faster than the poloidal contribution from parallel flow velocities [4]. Such measurements indicate that drifts could be the dominant transport mechanism in the stellarator SOL.
With respect to heat exhaust, a simplified model has been developed that provides a qualitative understanding of how the drifts impact the heatflux pattern observed on W7-X [5]. Additionally, advancements have been made in understanding the relationship between scrape-off layer (SOL) geometry and radiation behavior – finding new kinds of geometries associated with detachment instability [6,7]. Additionally, work is starting on first closed divertor designs for the stellarator island divertor [8]. Studies of how the open and closed divertor performance extrapolate to reactor-size are in progress. Finally, there have been recent advancements in comparisons of simulated heatfluxes of non-resonant divertor configurations with experiment [9].
[1] V. Perseo et al, this conference
[2] Y. Feng et al, Plasma Phys. Control. Fusion 53 (2011) 024009
[3] F. Sardei et al, J. Nuc. Mat. 241-243 (1997) 135-148
[4] C. Killer et al, Plasma Phys. Control. Fusion (2020)
[5] A. Kharwandikar, 29th IAEA Fusion Energy Conference (2023) London, UK
[6] V. R. Winters et al, Nucl. Fusion 64 (2024) 126047
[7] Y. Feng et al, Nucl. Fusion 64 (2024) 086027
[8] A. Menzel-Barbara et al, DPG Frühjahrstagung 2025, Göttingen, DE (Invited)
[9] N. Allen et al, APS Division of Plasma Physics 2024, Atlanta, GA (Poster)

Speaker: Victoria Winters (Max Planck Institute for Plasma Physics)

14:20 Heat Transport Widths and its Scaling in the W7-X Island Divertor

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 toroidally discrete divertor targets [2].
With the goal of evaluating the power exhaust performance of the island divertor, this contribution focuses on heat transport in the complex and inherently three-dimensional scrape-off layer (SOL) of W7-X. Drawing inspiration from established approaches in tokamak research — such as the use of the Eich function [3] to parameterize divertor heat loads — we present a novel framework to characterize the transport that is responsible for the two-dimensional heat flux patterns observed on the W7-X island divertor. By leveraging the spatial correspondence between strike line structures and SOL footprints on the divertor, the power exhaust can be systematically decomposed into three transport channels associated with distinct topological regions of the island SOL. Each channel is characterized by a representative length scale (width), which reflects the interplay between parallel and cross-field transport. Notably, the average power-channel width $\Lambda_W$ evaluated from the proposed scheme serves as a stellarator analogue to the well-known SOL heat flux width $\lambda_q$ in tokamaks.
These width parameters provide qualitative insights into transport processes dictating the divertor heat loads, enabling a systematic study of underlying mechanisms and allowing for consistent investigations of scaling with operational parameters. The framework has thus been applied to the experimental data from multiple W7-X campaigns [4]. Resulting empirical scalings of SOL transport widths with relevant plasma and operational parameters are presented and compared to simplified models. Finally, we highlight future prospects for the methodology, including cross-device comparisons of SOL transport, informing numerical models for 3D SOL, and standalone stellarator divertor heat load predictions.

[1] M. Endler et al, Fusion Engineering and Design 167 (2021) 112381
[2] Y. Feng et al, Nuclear Fusion 46.8 (2006) 807
[3] T. Eich et al, Phys. Rev. Lett. 107 (2011) 215001
[4] Y. Gao et al, Nuclear Fusion 59.6 (2019) 066007

Speaker: Mr Amit Kharwandikar (Max Planck Institute for Plasma Physics)

14:40 Spectroscopic investigations of island plasma parameters and impurity enrichment in the W7-X island divertor

The island divertor concept implemented at W7-X is currently one of the extensively investigated concepts for the power and particle exhaust for a future quasi isodynamic stellarator power plant. In the latest experimental campaign (OP2.3), W7-X demonstrated successful feedback control of radiative detachment via impurity seeding based on real-time measurements of the total radiated power via Bolometry. This capability marks an important step toward reactor-relevant power exhaust scenarios, where strong impurity radiation in the divertor is required to protect plasma-facing components. At the same time, future fusion reactors must limit the impurity content in the core plasma to sustain the fusion reaction. Impurity enrichment ($\mathrm{c}_\mathrm{imp,SOL}/\mathrm{c}_\mathrm{imp,core}$) is the performance parameter that characterizes the divertors ability to retain impurities and allow larger divertor concentrations without hampering core performance.
This contribution presents measurements and validation thereof of local plasma parameters in the W7-X divertor plasma based on line-ratio spectroscopy for Neon and Nitrogen. Results suggest that W7-X operates at relatively low Scrape Off Layer (SOL) electron densities ($1 - 5 \times 10^{19}~\mathrm{m}^{-3}$) compared to other devices (e.g., AUG, JET) that typically operate with electron densities $> 10^{20}~\mathrm{m}^{-3}$.
We show that transport cannot be neglected in the ionization balance in the W7-X island. The strong deviations from local equilibrium are likely driven by the low SOL densities and the open geometry of W7-X increasing among others processes the importance of neutral transport. For concentration estimates not only the local plasma parameters, but also the ionization balance is important. The inferred impurity concentrations in the divertor plasma are cross-validated using a newly installed time-of-flight (ToF) mass spectrometer in the sub-divertor volume part of the diagnostic residual gas analyzer (DRGA). To assess the enrichment parameter, the data is compared to core concentrations obtained via Charge Exchange Recombination Spectroscopy (CXRS) during diagnostic NBI phases.
We discuss two exemplary discharges seeded with Neon in the standard magnetic configuration. No significant enrichment is observed for Neon in detached conditions
($\mathrm{c}_\mathrm{imp,SOL}/\mathrm{c}_\mathrm{imp,core}$ ≈ 1). Ongoing work aims to extend the analysis to additional Neon cases and other seeded impurities such as Nitrogen and Argon. Notably, elevated edge electron densities diagnosed via Stark broadening and the presence of recombination radiation are observed in discharges with high core density and heating power. Such conditions could help to retain impurities in the SOL, limiting core contamination, and will therefore be investigated further.
Future experiments and divertor design optimizations (e.g., improved baffling to enhance neutral pressure) may offer pathways to achieve better impurity enrichment in stellarator island divertors. The present contribution provides a new toolset to study and quantify these effects in present day stellarators and beyond.

Speaker: Frederik Henke

15:00 Increased power operation with water-cooled divertors at Wendelstein 7-X

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 baffling.
During the initial experimental campaigns, the concept was tested using inertially cooled targets. Due to the open nature of their geometry, these experiments revealed low neutral pressures behind the divertors (well below 0.2 Pa). However, no significant issues were encountered in terms of heat load handling within the explored power and density range (≤ 6 MW and ≤ 10.5$ ⋅ $10$^{19}$ m$^{-3}$), for both attached and detached scenarios. In fact, the inertially cooled targets were robust and could reach high surface and bulk temperatures. The high material temperatures resulted in the release of a high number of particles, leading to power-starvation detachment through the radiation of intrinsic impurities (mostly carbon), with line-averaged densities well within the operational range for optimal electron cyclotron resonance heating (ECRH).
After two experimental campaigns, the targets were replaced with water-cooled ones, suitable for long pulse operation, but with a lower surface temperature limit. This hindered the ability to achieve high radiated power fractions ($f_{rad}$) with intrinsic impurities: higher line-averaged densities were required to reach the same level of $f_{rad}$ at the same power as with the uncooled divertor. The densities necessary for detached operation increased further when the power entering the scrape-off layer rose above 5 – 6 MW. Thanks to improvements made to the main heating systems (ECRH and neutral beam injection), input powers up to 13 MW were achieved, pushing the line-averaged density required for $f_{rad}$ > 0.8 closer to the limit of optimal ECRH absorption (2 ⋅ 10$^{20}$ m$^{-3}$).
Additional differences in divertor operation were observed in the high power experiments. The density profiles measured by multiple diagnostics (multi-purpose manipulator probes, Alkali beam, He beam) showed steeper radial gradients, both inside and outside the magnetic islands. Bolometer measurements revealed increased toroidal variation in plasma radiation at higher power levels, exacerbating the already existing discrepancy with EMC3-Eirene predictions. The high powers entering the scrape-off layer, now sustaining higher divertor densities, led to higher neutral pressures, stronger impurity flow velocities, and increased perpendicular transport. This change in transport resulted in heat loads being deposited on plasma-facing components that were not designed to withstand them. This prevented safe attached operation ($f_{rad}$ < 0.75) within the mid-density range (6 to 9 ⋅ 10$^{19}$ m$^{-3}$ line-averaged density) in some of the most commonly used W7-X magnetic configurations for heating powers above 4 – 5 MW, highlighting the need for reliable scenarios featuring seeded impurities (e.g. with a dedicated feedback controller).

Speaker: Valeria Perseo (Max-Planck-Institute für Plasmaphysik, Greifswald)

15:20 → 15:50 Coffee Break

15:50 → 17:15 Scrape-off-Layer and Divertor Physics

Convener: Marco Wischmeier (IPP Garching)

15:50 Long-pulse detached scenario by feedback N2 seeding on EAST

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 the feedback control of the divertor electron temperature (T_{e,div}) of 5eV and extends the detachment control time to about 60s, which is the longest record of the active detached plasma by the feedback control on EAST. The N_{2} seeding reduces the divertor heat flux, supressing the tungsten (W) sputtering and generates the local strong radiation near X-point. EAST also uses the clean discharge to remove the wall- retained N particles. The heating power of ECRH (P_{ECRH}) plays an important role in the N-removing, applying the clean discharge in the middle of a series of discharges is an effective way to apply the N_{2} seeding in the long-pulse operation of the detachment phase.

Speaker: Kai Wu (Institute of Plasma Physics, Hefei institutes of physical sciences, Chinese academy of sciences)

16:10 The “Chimney” divertor: A closed divertor with mid-leg pumping for core-edge integration in DIII-D

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 new paradigm for addressing the integrated tokamak exhaust and performance challenge with reduced requirements for divertor volume and radiating impurity injection.

This “Chimney” design employs a closed divertor structure along an extended outer leg with a pump duct opening positioned vertically upstream from the target on the common flux side. A 4.5 cm deep volume at the end of the divertor slot acts as a neutral reservoir surrounding the outer target to dissipate plasma thermal exhaust via neutral friction, charge exchange and recombination downstream of where neutral particles are removed by the pump. While positioning the pump upstream of the target reduces the fraction of fast reflected particles captured on their first flight by the pump opening, a high fraction of thermalized neutral particles is still removed by the pump before they reach the main chamber where they can ionize on closed flux surfaces and increase the core plasma density. The hypothesis is that this also stops the neutral dissipation processes from propagating further upstream, thereby stabilizing the detachment front. The width of the divertor slot opening is designed to accept the majority of exhausted plasma flux while remaining narrow enough to minimize leakage of neutral particles back to the main chamber, with room for experimental testing and optimization.

Behavior of the chimney pump design is predicted with 2D boundary fluid plasma and neutral modeling using both UEDGE [1] and SOLPS-ITER [2]. These simulations predict that positioning the pump duct upstream from the target in the divertor slot reduces the upstream separatrix density at the onset of detachment. They also find that the location of the 10 eV front along the outer divertor leg is stabilized poloidally near the pump duct opening over an experimentally attractive range of injected power and D2 gas puffing. Plasma control analysis finds a realizable equilibrium with a $\leq$ 44 cm outer leg poloidal length using 1.8 MA of plasma current, resulting in an edge safety factor q$_{95}$=3.3. This equilibrium is calculated to have sufficient flexibility to scan the length and position of the outer leg.

A comprehensive set of diagnostics is planned to be installed to monitor plasma conditions and dissipation processes in the new divertor. Tiles will be instrumented with 44 in-situ Langmuir probes, 22 surface-eroding thermocouples, 6 ASDEX-style neutral pressure gauges, and 2 in-situ Penning gauges. New Thomson scattering measurement positions will enable diagnosis of Te near the X-point, and new in-vessel fiber optic views will complement an existing tangentially-viewing camera to diagnose impurity radiation profiles along the outer leg.

[1] A. Holm et al., Nuclear Materials and Energy 41 (2024), 101782.
[2] J. H. Yu et al., Nuclear Materials and Energy 41 (2024) 101826.

This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC52-07NA27344, DE-NA0003525, and DE-SC0023378.

Speaker: Robert Wilcox (Oak Ridge National Laboratory)

16:30 Discussion

17:15 → 18:15 Programme Commitee Meeting [Closed Session]

19:00 → 22:00 Conference Dinner

Friday 31 October

09:00 → 10:25 Towards Integrated Scenarios for Exhaust

Convener: Matthijs van Berkel (DIFFER)

09:00 Topical session 5 - Towards Integrated Scenarios for Exhaust

Speaker: Matthijs van Berkel (DIFFER)

09:05 The X-Point radiating regime at JET in D and DT plasmas with mixed impurities

Power exhaust is a crucial issue for future fusion reactors. To reach the required 95% dissipation of the exhaust power, impurities must be injected into the plasma. With strong impurity seeding the radiation concentrates in a small region inside the confined plasma, forming the X-point radiator (XPR). The XPR is observed in almost all currently operating tokamaks and was, in JET, first observed in 2016. In the recent campaigns, which culminated in the last DT campaign of JET (DTE3) and the subsequent shutdown of the machine, the XPR was investigated in detail.

Several seed impurities were injected in order to trigger an XPR, such as nitrogen, neon, argon, and combinations thereof. With pure neon or argon seeding the plasma exhibits dithering between H- and L-mode, even at heating powers of up to 26 MW. The dithering is suppressed and the plasma stays in H-mode when combining two impurities or with pure N2 seeding. As the use of N2 was not permitted in the DT campaign, the mixture of Ar and Ne performed best while still being compatible with DT operation. The impact of the seed impurities used and the benefits of the mixed seeding are presented. With the help of SOLPS modelling the effect of the impurities is analysed in terms of their impact on access to the XPR regime, radiative capability, H-mode stability, and plasma performance.
For the first time at JET, a movement of the XPR inside the confined region was observed (see Figure 1). This movement was tracked by the horizontal bolometer camera, using the algorithm developed at AUG, with the XPR moving mainly between lines of sight 3 and 4. The XPR location is detected to a sub-channel accuracy of 4 mm (with a channel spacing of about 8 cm) and can be provided in real time to the control system. A PI controller is implemented using Ar seeding as actuator (while the Ne injection is pre-programmed). The controller gains were optimized using system identification experiments. The reaction time of the XPR location to a change in the seeding rate is in the range of 1s, which presents a significant restriction for the control of the XPR. However, external perturbations, such as drops in heating power or pellet injection, are first buffered by a movement of the XPR, and can then be effectively counteracted by the slower control. These dynamics will be compared to the faster reaction times of the XPR in AUG.
The active control of the XPR helped to efficiently establish the same power exhaust conditions when moving from pure D plasmas to DT plasmas. The overall performance of the scenario, though initially low without seeding (H98≈0.65), does not decrease with the impurity injection, and increases slightly for the seeded and unseeded case when going to DT. Notably, the edge kinetic profiles are not observed to be affected by the strong seeding, while ELMs become fully mitigated.
The successful demonstration of the XPR in D and DT plasmas at JET, combined with its stable operation, makes a strong case for incorporating XPRs into future fusion devices. This scenario meets several requirements of a reactor, including high power dissipation, control of full detachment, and effective ELM heat load mitigation, though not at highest confinement yet.

Speaker: Matthias Bernert (Max-Planck-Institut für Plasmaphysik)

09:45 Exhaust control in alternative divertors for transient heat load management

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 divertor conditions is challenging as the power fluctuations can be too fast for actuators to respond to. Alternative Divertor Configurations (ADCs) offer a potential solution to this problem due to their superior performance compared to conventional divertors [1,2].

Our work on MAST-U successfully demonstrates power exhaust control in ADCs (Figure 1), representing the first such demonstration beyond conventional divertors [3,4]. This is achieved through novel sensor techniques, enabling control of the detached plasma in between the target and the X-point in real-time using $\mathrm{D}_2$ Fulcher emission measurements. Detachment control was not possible in conventional divertor scenarios on MAST-U, as their divertor state was too sensitive to perturbations, giving gas actuators insufficient time to actuate.

We demonstrate that ADCs can tackle key risks and uncertainties for fusion energy: 1) their highly reduced sensitivity to perturbations enables active exhaust control in otherwise unfeasible situations and facilitates 2) an increased passive absorption of fast transients which would otherwise damage the divertor; furthermore, we observe 3) a strong isolation of each divertor from other reactor regions through tight baffling which prevents divertor neutrals from spreading into the main chamber and effecting the core plasma.

This divertor isolation is evidenced in Figure 1d, where exhaust control using a divertor fueling valve did not influence core plasma conditions, contrary to midplane valve exhaust control experiments [5]. This enables near-independent control of the divertors and core plasma which, although highly beneficial to all reactor concepts, is essential to compensate the asymmetric power transients expected in reactors with Double-Null divertor configurations.

The recent introduction of a cryopump in MAST-U further improves control capabilities by allowing the detachment front to be moved downwards, towards more attached conditions. This marks a significant improvement over previous experiments without cryopumping where a buildup of neutral pressure prevented the transition to more attached conditions [5]. This highlights the importance of adequate pumping for exhaust control.

In summary, our results demonstrate the practical advantages of ADCs for effective heat load management in fusion power reactors.

Figure 1: Exhaust control in the MAST-U Super-X divertor using cryopumping and divertor fueling. (a) $\mathrm{D}_2$ Fulcher-band filtered image of the lower divertor, showing the tracking area (green box) [6], maximum intensity (black circle), detected divertor leg (orange dots), and the detected emission front (red cross). (b) Corresponding divertor cross-section with magnetic divertor topology (black), detected divertor leg (orange dots), and the detected emission front (red cross). The red arrow indicates the distance-to-target measurement Ltar. (c) Time evolution of the emission front position (Ltar) compared to the reference signal (dashed). Cryopumping also allows the front to be moved down, closer to the target, contrary to non-cryopumped experiments [5]. (d) Line-integrated core density, showing no response to divertor actuation, contrary to midplane fueling experiments [5]. (e) Gas flow request to the lower divertor valve by the exhaust controller.

[1] Verhaegh, K. et al. Divertor shaping with neutral baffling as a solution to the tokamak power exhaust challenge. Nature Commun Phys 8, 215 (2025). https://doi.org/10.1038/s42005-025-02121-[2] Theiler, C. et al. Results from recent detachment experiments in alternative divertor configurations on TCV. Nucl. Fusion 57, 072008 (2017). https://iopscience.iop.org/article/10.1088/1741-4326/aa5fb7
[3] Ravensbergen, T. et al. Real-time feedback control of the impurity emission front in tokamak divertor plasmas. Nature Communications 12, 1105 (2021) https://www.nature.com/articles/s41467-021-21268-3
[4] Koenders, J.T.W. et al. Model-based impurity emission front control using deuterium fueling and nitrogen seeding in TCV. Nuclear Fusion 63, 026006 (2023) https://iopscience.iop.org/article/10.1088/1741-4326/aca620
[5] Kool, B. et al. First demonstration of Super-X divertor exhaust control for transient heat load management in compact fusion reactors. Nature energy accepted preprint available at https://www.researchsquare.com/article/rs-5059325/v1
[6] Ravensbergen, T. et al. Development of a real-time algorithm for detection of the divertor detachment radiation front using multi-spectral imaging. Nuclear Fusion 60 (2020). https://iopscience.iop.org/article/10.1088/1741-4326/ab8183

Speaker: Bob Kool (DIFFER/Eindhoven University of Technology)

10:05 Core-edge Integrated Simulations of Ne-seeded JET-ITER Baseline Scenarios

The Ne-seeded JET-ITER baseline is a robust scenario that can achieve simultaneous partial divertor detachment and high-confinement (H$_{98}$ > 0.85-1.0) with small or no ELMs at JET [1, 2, 3]. In these highly fuelled, high-triangularity scenarios with vertical target divertor configuration, Ne-seeding opens access to reduced pedestal top electron density, $n_\text{e,PED}$, with low pedestal collisionality and improved overall confinement, primarily driven by pedestal pressure increase relative to the unseeded conditions. Simultaneously, ELMs are reduced in size and eventually disappear with increasing neon content. This contribution reviews the recent activities in core-edge integrated simulations and analysis of these scenarios with standalone SOL transport simulations with EDGE2D-EIRENE, SOLEDGE3X, and SOLPS-ITER supporting integrated JINTRAC-COCONUT analysis. The chosen approach is such that SOLEDGE3X and SOLPS-ITER pursue detailed boundary model validation, while EDGE2D-EIRENE is used to bridge the integration gap to the JINTRAC-COCONUT simulations. A key goal for the core-edge integrated simulation workflow is to provide physics insight to the main drivers responsible for the $n_\text{e,PED}$ reduction at JET as well as to address how these drivers impact the overall divertor design and extrapolation of exhaust-integrated scenarios to the scale of reactors.

The $n_\text{e,PED}$ reduction is hypothesized to be driven by a combination of (1) increase in pedestal particle transport, (2) reduction of ionization sources inside the pedestal, $S_\text{IZ,PED}$, and (3) reduction of separatrix electron density, $n_\text{e,SEP}$, due to the onset of detachment and starvation of the divertor plasma from power to ionize neutrals. While gyrokinetic simulations of the pedestal plasmas have not yet identified a turbulent flux compatible with the observed reduction of $n_\text{e,PED}$, the SOL simulations indicate the presence of the other two mechanisms (2, 3) with their magnitude depending on user-assumptions of the uncertain code parameters. For example, EDGE2D-EIRENE simulations predict a ~x4 reduction of $S_\text{IZ,PED}$ and about a 25% reduction of $n_\text{e,SEP}$ with Ne-seeding. These observations are associated with a x2-3 reduction in the neutral pressure in the vacuum region surrounding the plasma in the simulations, consistent with experimentally observed ~x3 reduction in the horizontally line-integrated $D_\alpha$-light at the mid-plane level, including sub-traction of reflections according to the method detailed in [4]. Both results give an indication that reduction in $S_\text{IZ,PED}$ is likely to play a role in the $n_\text{e,PED}$ reduction. The exact balance between the three drivers can only be narrowed by the detailed validation effort, including extended grids and drifts with SOLEDGE3X and SOLPS-ITER. The boundary simulations are now used to guide the JINTRAC-COCONUT analysis for a calibrated core-edge simulation effort to proceed towards an integrated analysis of the exhaust-integrated pedestal at Ne-seeded JET-ITER baseline.
[1] C. Giroud, et al. IAEA-FEC 2025.
[2] C. Giroud, et al. PSI 2024.
[3] C. Giroud, et al. IAEA-FEC 2021.
[4] L. Horvath, et al. Plasma Phys. Control. Fusion 65 (2023) 044003.

Speaker: Aaro Järvinen (VTT)

10:25 → 10:55 Coffee Break

10:55 → 12:40 Towards Integrated Scenarios for Exhaust

Convener: Matthijs van Berkel (DIFFER)

10:55 Advances in Neutral Particle Control in the Closed Helical Divertor of LHD

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 of the study presents the conceptual evolution, engineering realization, and experimental validation of the CHD. The integration of cryo-sorption and non-evaporable getter (NEG) pumps significantly improved neutral particle exhaust and reduced wall recycling, enabling flexible density control and sustained long-pulse operation. Fast ion gauges and EMC3-EIRENE simulations revealed more than a tenfold increase in divertor pressure with minimal impact on core plasma parameters.
The second part focuses on recent observations of ultra-high neutral pressures (up to 2.4 Pa) in inward-shifted magnetic configurations (Rax = 3.55 m). This regime is attributed to a newly identified near-wall condensation phenomenon, where recombination in a dense, low-temperature divertor plasma leads to a secondary neutral source. The resulting localized neutral buildup enhances divertor compression and leads to radiation detachment, observed via bolometry, spectroscopy, and high-speed imaging.
Together, these results demonstrate the viability of advanced divertor designs and localized neutral source phenomena as effective tools for achieving efficient particle exhaust and plasma edge control in stellarator reactors. These insights contribute to the development of reactor-relevant divertor concepts for future steady-state fusion power plants.

Speaker: Dr Gen Motojima (National Institute for Fusion Science)

11:15 Progress towards Integrated Tokamak Scenarios for Exhaust: experiments and new self-consistent core-edge modeling framework

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 available for extrapolation to future power plants.
Highly radiating plasmas in negative triangularity (NT) have been demonstrated through the use of reactor-relevant seeding gases, i.e., neon, argon, and krypton, as extrinsic impurities featuring simultaneously high performance (𝛽N >2), divertor heat flux reduction and intrinsically no ELMs [1]. We present a comprehensive core and divertor modeling which highlights the physics mechanisms leading to confinement improvements and simultaneous reduction of the divertor heat flux when mantle radiation is integrated with the NT configuration. Seeding with Kr and Ar lead to a reduction of the parallel heat flux at the divertor entrance compared to N and Ne effectively alleviating the power exhaust by reducing PSOL, which is one of the main advantages of working in NT configuration. Higher impurity compression for Ar and Kr is found compared to N and Ne. The high-Z impurities provide volumetric dissipation at lower concentrations at the separatrix up to ~90% in agreement with experimental estimates for fuel dilution. Thus, the same divertor conditions can be obtained with a reduced impurity concentration at the separatrix, which is important for integrated scenarios for exhaust. The results presented here support that there is a path to highly radiating, high performance NT plasmas with low PSOL and no ELMs which all enable a stable plasma material interface. Additionally, for future reactors, a NT shape naturally puts the divertor at large major radius, which means a larger wetted area for power loads is available with a great potential to improve core–edge integration, easing the divertor installation, remote access and maintenance technology.
Core-edge integrated simulations with the new developed SICAS framework (SOLPS-ITER Coupled to ASTRA-STRAHL) [2] provides self-consistent background plasma and impurity transport from the divertor to the core with good agreements with experimental data. This tool opens new possibilities in integrated modeling of fusion devices for the interpretation of current experiments, prediction for ITER as well as reactor design. Application of SICAS on other impurity seeding scenarios including closed divertors will be also discussed. By capturing the complex interplay between impurity transport, core confinement, and edge dissipation, SICAS provides a physics-based foundation for designing new integrated scenarios for exhaust.

[1] L Casali et al 2025 Plasma Phys. Control. Fusion 67 025007, [2] A. Welsh et al 2025 Nucl. Fusion 65 044002
Work supported by the U.S. Department of Energy, under Award(s) DE-SC0023100, NRC 31310022M0014, DE-FC0204ER54698, DE-SC0022270, DE-AC52-07NA27344, DEFG02-08ER54999.

Speaker: Livia Casali (University of Tennessee Knoxville)

11:35 Simulation study of the grassy ELM cycles within Edge Plasma Coupling Simulation framework

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 simulations are indispensable for both understanding the edge plasma physics and predicting the edge plasma behavior. However, the cross-field transport coefficients adopted in the transport codes such as SOLPS-ITER, UEDGE, EDGE2D-EIRENE and SOLEDGE2D-EIRENE are usually given empirically or by fitting experiments. On the other hand, although the turbulent cross-field transport of the edge plasma can be simulated by lots of turbulence codes such as BOUT++ and JOREK, it is hard to achieve a self-consistent simulation of the edge plasma transport due to the large gap between the turbulence and transport time scales. One possible way to realize a self-consistent edge plasma simulation is the coupling simulation by the transport and turbulence codes [1, 2].
In our recent work [3], a simulation framework called EPCS (Edge Plasma Coupling Simulation) is developed for the purpose to implement the self-consistent turbulence-transport coupling simulation of the edge plasma automatically and efficiently. Based on a steady-state coupling simulation workflow, the edge plasma is simulated by iterations of turbulence code BOUT++ [4] and transport code SOLPS-ITER [5], and the converged plasma profiles are consistent with EAST experiments (edge-localized-mode-free stage) at both upstream and divertor target.
Grassy ELM regime is considered as a possible edge plasma solution due to the desirable features including significant reduction of transient divertor heat fluxes and quasi-continuous particle and power exhaust [6]. To investigate the formation mechanisms of the grassy ELMs, a time-dependent coupling simulation workflow is developed and the grassy ELM cycles are simulated within EPCS. The simulated ELM cycles show a counterclockwise trajectory in the peeling-ballooning diagram with an ELM frequency ~2 kHz, which is consistent with EAST experiment. According to the simulation result, the particle transport is strong at the pedestal foot rather than the region with steep pressure gradient, which can be explained by the influence on the profile of radial electric field shear due to the scrape-off layer plasma.

Reference
[1] T. D. Rognlien, et al., Contrib. Plasma Phys. 44 (2004) 188–193.
[2] D. R. Zhang et al., Nucl. Fusion 60 (2020) 106015.
[3] T. Y. Liu et al., Plasma Phys. Control. Fusion 67 (2025) 055004.
[4] X. Q. Xu et al., Phys. Plasmas 7 (2000) 1951–1958.
[5] S. Wiesen et al., J. Nucl. Mater. 463 (2015) 480–484.
[6] G. S. Xu et al., Phys. Rev. Lett. 122 (2019) 255001.

Speaker: Prof. Minyou Ye (University of Science and Technology of China)

11:55 Discussion

12:40 → 13:05 Closing

13:05 → 14:30 Lunch