4th Technical Meeting on Divertor Concepts
Board Room A
IAEA Headquarters
KEY DEADLINES
30 June 2022 Deadline for submission of Participation Form (Form A), and Grant Application Form (Form C) (if applicable) through the official channels
11 July 2022 Deadline for submission of abstracts through IAEA-INDICO for regular contributions
31 August 2022 Notification of acceptance of abstracts and of assigned awards
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.
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Opening and Introduction Board Room A
Board Room A
IAEA Headquarters
Conveners: Marco Wischmeier (IPP Garching), Matteo Barbarino (International Atomic Energy Agency)-
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IAEA IntroductionSpeaker: Matteo Barbarino (International Atomic Energy Agency)
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Chair IntroductionSpeaker: Marco Wischmeier (IPP Garching)
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DEMOs and Next Step Facilities Board Room A
Board Room A
IAEA Headquarters
Conveners: James Harrison (United Kingdom Atomic Energy Authority), Nobuyuki Asakura (National Institutes for Quantum and Radiological Science and Technology (QST))-
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The EU-DEMO Exhaust Modelling Roadmap – Numerical Implementation and Methods
A simple extrapolation of the ITER Q=10 H-mode divertor target heat-load specification assuming mitigated type-I ELMs is not sufficient for the exhaust concept required in EU-DEMO provided the increased power level of Pedge = 300MW entering the edge region. The combination of an anticipated reduced power fall-off length of only a few mm in the SOL at a plasma current Ip=20MA whilst keeping the increase of the size of the device at a moderate level to optimize for cost of magnetic field (scaling with major radius, R_DEMO ~ 1.5 R_ITER) requires an integrated core-edge scenario with a tailored impurity mix inducing an energy dissipation fraction of up to 95% in the edge (of which 30% must occur in the confined region by line radiation). In recent experiments it has been demonstrated that at such high levels of radiation (e.g. located close or at the X-Point in single-null configurations) a transition into a small/no-ELM regime is observed coinciding with controllable strong detachment in the divertor. Modelling of an X-Point radiating regime and the validation of a suitable model is an ongoing task. Improved core physics scenarios for EU-DEMO (not necessarily employing an X-Point radiating regime) require upgrades for the pedestal transport model, or a lifting of other geometrical constraints. For example, the inclusion of a secondary X-Point in the upper-plane allowing strongly shaped plasmas at high beta may cause an additional source of impurities from wall erosion by energetic particles in remote areas and requires further numerical assessments with plasma-wall interaction codes.
The conceptual design phase for EU-DEMO implies a revision of the exhaust modelling roadmap until 2024. For the purpose of identifying a controllable exhaust scenario to be employed in EU-DEMO, the required physics foundation of candidate regimes must be re-assessed by using validated numerical tools. A new baseline reference exhaust model setup is currently being established in the DEMO central team (DCT) in Europe. This contribution summarizes the key aspects of the proposed complete SOLPS-ITER EU-DEMO physics model, including: fluid drifts and neutral kinetics, allowing for charge exchange for impurities enhancing non-coronal radiation levels in the edge. A multi-strand approach employed by the DCT with different levels of fidelity is presented, that seems advantageous to complement the final model to be composed for exploring the boundaries of the EU-DEMO exhaust operational window. Recently, new developments have started of an extension to the SOLPS-ITER code for automatic optimization of the DEMO divertor shape and a status of this work is presented. Finally, an outlook based on very recent activities on the development of fast exhaust models employing advanced machine learning will be given. Employment of such fast pre-trained models seem to be promising for fast integrated scoping studies through systems design codes and future plasma control schemes.
Speaker: Dr Sven Wiesen (Forschungszentrum Jülich GmbH, Institut für Energie-und Klimaforschung – Plasmaphysik, 52425 Jülich, Germany) -
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Divertor concept development for the W7-X stellarator experiment
The Wendelstein 7-X (W7-X) is an advanced stellarator device operated in Greifswald, Germany, to provide the proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high-performance, steady-state HELIAS operation. During the experimental programs, starting with the first operation phase OP1.1 (Dec. 2015 - March 2016) up to OP1.2a (Sept. 2017 - Dec. 2017) and OP1.2b (July 2018 - Oct. 2018), the energy input and plasma performance as well as the heat and particle loads on the in-vessel components were continuously increased. In these first OPs, the performance was limited because most of the in vessel components and in particular the divertor made of fine graphite elements was not water-cooled. During the experiments, high heat loads on the in-vessel components have been observed, which exceeded the specified limits under certain conditions, thus limiting the operation. In particular, significant baffle overload was detected in OP1.2 in the high mirror magnetic field configuration - the only configuration that meets all optimization criteria of W7-X. In addition to the baffle overload, an extension of the main strike-line to the middle divertor was observed in many configurations, resulting in high heat loads - larger than acceptable for the water-cooled CFC high-heat-flux divertor in the upcoming experiment campaign OP2.1 (start Nov. 2022). The installation of the water cooled divertor for OP2, which has the same geometry as the OP1.2 divertor, together with the cryo-vacuum pump system, has been completed in 2021. The purpose is to reach the defined goal of W7-X: 30 min long pulse operation with 10 MW plasma heating.
There is a general agreement within the W7-X project that the transition to reactor relevant materials for the plasma-facing components (PFCs) is an important and necessary step. The envisaged transition to an all-metallic PFC device requires a new design of the divertor components. The development of target elements with tungsten-based armor material designed to remove up to 10 MW/m2 in steady-state operation was started in early 2021 in the framework of EUROfusion.
Appropriate design changes taking into account various technical constraints need to be developed to optimize the current divertor geometry; a thorough evaluation and validation of the applied simulations against the experimental findings, as partially mentioned above, is being performed. The final goals will be the reduction of peak heat loads as well as high gas exhaust by divertor geometry and plasma scenario optimization.
This contribution presents first results of modeling activities for a new W7-X divertor using fast running tools such as the recently developed EMC3-lite for heat load analysis and an ANSYS application for exhaust studies to find favorable design changes, which are then verified with the comprehensive EMC3/Eirene and DIVGAS codes. These modeling activities are supported by the development of efficient engineering tools in a CATIA environment that process the complex 3D W7-X design data at different levels of sophistication to promote an efficient interchange with the physics-based codes. The impact of specific geometry modifications on divertor performance is reported.Speaker: Dirk Naujoks (IPP Greifswald) -
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Utilization of SPARC to investigate divertor solutions for fusion pilot-plants
SPARC is a compact, high-field short pulse ICRF heated tokamak ($B_{0}=12.2$ T, $R_{0}=1.85$ m, $\tau_{flattop}=10$ s, $P_{rf}=25$ MW) with a close-fitting tungsten first wall designed to achieve its mission goal of $Q_{fus}>2$ with significant margin. Construction has begun at a new site in Devens, MA and operations are scheduled to begin in mid-2025. Although the baseline design relies on strike point sweeping, divertor shaping, and inertially cooled divertor targets to facilitate attainment of this primary mission at lowest risk, the device is able to access reactor relevant pedestal and divertor parameters with unmitigated parallel heat fluxes entering the divertor of $\sim10$ GW/m$^2$, pedestal temperatures of $\sim3$ keV and separatrix densities $\sim10^{20}$ m$^{-3}$ ($n_{Greenwald}\sim8×10^{20}$ m$^{-3}$). Therefore, in addition to its fusion power goals, SPARC will also be a test bed for exploring the key areas of tokamak physics necessary to design an ARC-class fusion pilot plant. An anticipated focus of experiments will be on identifying what ranges in $H_{98}$ and $Q_{fus}$ are compatible with highly dissipative, low steady state erosion divertor scenarios, that also ensures neutral pumping levels that would be sufficient for helium ash removal.
The inclusion in the SPARC design of a long-leg divertor, magnetic topology flexibility, multiple gas injection locations (fuel, impurities) and neutral pump actuators will facilitate this goal by assessing variations in divertor geometries and operation scenarios being consider for an ARC-class device. The divertor is up-down symmetric, able to achieve standard horizontal- and vertical-target-like neutral recycling conditions on both the inner and outer divertors. The outer divertors also feature highly baffled long legs ($L_{pol}\sim0.5$ m) with coil control to also enable secondary X-points in the divertor volume at the end of the outer divertor legs to form a X-point target divertor magnetic geometry. Flexibility in gas injection has been included in the design with both main ion and impurity gas seeding nozzles located in the inner and outer divertors as well as at the primary X-point. Each of these injection locations is capable of independent injection control using up to 3 simultaneous gas mixtures. In addition, the divertor neutral pump inlet is located at the end of the outer divertor designed with an adjustable pumping speed of up to $\sim 20$ m$^3$/s, though it is not designed to pump helium.
However, due to the closed divertor shape and the need for neutron tolerant systems, diagnostic access will be limited. Hence, a workflow for boundary plasma simulations is being developed both as a means of interpreting experimental observations as well as informing experimental planning. A companion contribution (“SPARC Diagnostics for Use in Plasma Control and Divertor Physics Studies” – M.L. Reinke, et al.) will focus on the SPARC divertor diagnostic set. SPARC will provide near-term, invaluable experience with what it takes to diagnose and control the divertor plasma under reactor conditions with limited diagnostic access.Speaker: Adam Kuang (MIT Plasma Science and Fusion Center)
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10:40
Break Board Room A
Board Room A
IAEA Headquarters
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DEMOs and Next Step Facilities Board Room A
Board Room A
IAEA Headquarters
Conveners: James Harrison (United Kingdom Atomic Energy Authority), Nobuyuki Asakura (National Institutes for Quantum and Radiological Science and Technology (QST))-
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Physics basis and design of tungsten divertor for CFETR
Significant advances have been made on the physics design of full tungsten divertor and the edge modeling for Chinese Fusion Engineering Testing Reactor (CFETR). CFETR is proposed by Chinese fusion community to bridge the gap between ITER and DEMO with fusion power up to GW level [1]. One of the key challenges is that the divertor solution for CFETR must meet requirements beyond that of ITER due to higher duty cycle and higher power across separatrix per unit major radius. Taking into account the engineering requirements, a conventional full tungsten (W) divertor with long divertor leg length and V-shape corner structure has been proposed [2]. The SOLPS-ITER code package with full drifts and currents are employed to evaluate the divertor performance of two candidate radiation impurity species, argon (Ar) and neon (Ne), with two divertor geometries (baseline and long leg divertor) [3]. The self-consistent core-edge integrated code COREDIV coupling with the iteratively calculations of core plasma within the OMFIT framework are used for core-edge integration simulations, which helps to set requirements for SOLPS edge modelling [4], such as the separatrix plasma density, impurity concentration etc. The modeling results show clearly that increasing the seeding rate of Ar or Ne can reduce the target electron temperature and heat flux, which can be reduced further by higher D2 injection rate. Similar core plasma and divertor conditions, as well as radiated power fraction, can be achieved with 2-3 times less Ar seeding rate than the Ne seeding. In spite of better argon radiation efficiency, no significant increase of core radiation with argon seeding is observed compared to neon seeding, which is due to better argon compression near the divertors caused by its smaller first ionization potential. Longer divertor leg length has a distinct advantage on radiation losses, which has been adopted for the current engineering design. Based on the background plasma from the SOLPS modeling, W erosion and edge transport has been estimated by using the DIVIMP code. The W sputtering is mainly contributed by impurity ions at the far SOL region due to high electron and ion temperature there, but the W net erosion rates at both divertor targets are below the target lifetime requirements for CFETR operation. Shaping of W plasma-facing components (PFCs) is designed to avoid leading edges due to misalignment, which can increase the stationary heat flux by ~49%. Transient heat flux has been calculated using the BOUT++ simulations, which shows a grassy ELMy characteristic for hybrid scenario. The small ELM size will not melt the W PFCs even the shaping effects are considered. The possible divertor solution was obtained for CFETR which can meet the physics requirements on target heat flux, target lifetime and core compatibility.
[1] G. Zhuang, et al., Nuclear Fusion 59 (2019) 112010
[2] X. J. Liu et al., Physics of Plasmas 27 (2020) 092508
[3] H. Si et al., Nuclear Fusion 62 (2022) 026031
[4] H. Xie et al., Nuclear Fusion 60 (2020) 046022Speaker: Dr Rui Ding (Institute of Plasma Physics, Chinese Academy of Sciences) -
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Physics drivers of the STEP divertor concept design
The STEP programme aims to demonstrate viability of a GW-scale spherical tokamak reactor, with parameters around: geometric axis 3.6m, aspect ratio 1.8, elongation 2.9, plasma current 20MA, on-axis magnetic field 3.2T, core radiated power 350MW, with steady state power crossing the separatrix 150MW. Here we give a physics overview of the studies driving forward the STEP divertor concept design, focussing on operational phases during current ramp-up and flat-top.
The inner divertor in a spherical tokamak is challenging, with total flux compression along the leg and small strike point major radius suggesting unmitigated surface power loads of GW/m$^2$. The down-selected operational point thus focused around a double-null configuration, with surface area and poloidal variation of transport weighting the core power outflux towards the outboard. The outer divertor leg was extended to large major radius to increase the target wetted area, with unmitigated loads up to ~200MW/m$^2$, managed by seeding. Our recent studies indicate how synergy between connection length and flux expansion can be used to improve the detachment threshold and window, as well as front location sensitivity. Advantages of double-null geometry will only be realised with vertical position control precision to 1-2 millimetres, based on anticipated heat flux decay widths, so we have examined the impact of separatrix disconnection, dR$_{\rm sep}$, on target fluxes. Up-down power sharing was found to be consistent with previous experimental modelling [Brunner et al (2018) Nucl. Fusion 58 076010], but the strong flux compression from outboard midplane to inboard target contributed to weak sensitivity of in-out power sharing, with the impact of drifts on this conclusion under investigation. This motivates testing of dynamic double-null (full power alternating from upper to lower divertor) for practical vertical stability control, with modelling indicating strong frequency and amplitude dependence of divertor component fatigue by temperature cycling. Partial detachment was recovered in a vertical inner target configuration, with high far-SOL temperatures driving unacceptable erosion. Full detachment required direct extrinsic impurity seeding or the introduction of an advanced divertor geometry, with the former giving unacceptably high core fuel dilution.
A toolkit has been developed which allows optimisation of divertor magnetic geometries in free boundary magnetic equilibria, in combination with compatible first wall geometries, while holding constraints within allowed margins. Constraints include core profiles, informed by stability and transport analysis, and field coil positions, defined by neutron shielding and current limitations, while divertor performance measures, such as connection length and poloidal flux expansion, are returned.
Performant geometries are passed to SOLPS-ITER, to assess trends in power handling ability.
The detached operating space is identified in terms of deuterium puffing and argon seeding, while minimising gas throughput, and maintaining low upstream density and impurity concentration. The inner divertor now focuses on a design approaching an X-divertor, finding substantial improvements in performance allowing reduced seeding levels, with ionisation profiles capturing the benefits of angled targets. The effect of pump location and inner target geometry on helium pumping efficiency is being optimised. The potential impact of gas puff location and pumping speed are also studied.Speaker: Sarah Newton (Culham Centre for Fusion Energy) -
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Simulation studies of He and particle exhaust in detached divertor for JA DEMO design
SONIC divertor code enables simultaneous calculations of seeding impurity (Ar) and fusion product (He ash) transport. He exhaust has been investigated in JA DEMO, where exhaust power ($P_{out}$ = 250 MW), ion flux ($\Gamma_{out}^{D}$ = $\rm 1x10^{22} s^{-1}$) and He ion flux ($\Gamma_{out}^{He}$ =$\rm 5.3x10^{20} s^{-1}$, corresponding to 1.5 GW fusion power) were given at the core-edge boundary. Plasma diffusion coefficients of $\chi$ = 1 $\rm m^{2}s^{-1}$ and $D$ (plasma and impurity ions) = 0.3 $\rm m^{2}s^{-1}$ were the same as “standard” values in the ITER simulation. Peak heat loads at the inner and outer divertor targets were reduced less than 10 $\rm MWm^{-2}$ for a reference series of the radiation fraction in the SOL and divertor, i.e. $f_{rad}^{div}$ = $(P_{rad}^{sol}+P_{rad}^{div})/P_{sep}$ $\sim$0.8. He concentration ($c_{He}^{edge}$ = $n_{He}/n_{i}$) averaged at the plasma edge near the midplane ($r^{mid}/a_{p}$ = 0.96-0.98) was evaluated in the detached divertor condition; fully and partially detachment in the inner and outer divertors, respectively. $c_{He}^{edge}$ was reduced from 6 $\%$ to 4 $\%$ with increasing $n_{e}^{sep}$ from $\rm 1.8x10^{19}$ to $\rm 2.3x10^{19} m^{-3}$ by $\rm D_{2}$ gas puff (keeping the same $f_{rad}^{div}$ $\sim$0.8 by reducing Ar seeding rate), while the partial detachment was extended. In the divertor, in-out asymmetry of $c_{He}^{div}$ was seen (2-3 times); $c_{He}^{div}$ in the upstream of the inner divertor was enhanced to larger than 10 $\%$, maybe caused by large thermal force (parallel ion temperature gradient) on He ions in the fully detached condition.The in-out asymmetry were reduced near the separatrix of the main SOL.
Influences of reducing $\chi$ and $D$ on the He exhaust were investigated ($\chi$ = 0.5, $D$ = 0.15 $\rm m^{2}s^{-1}$), compared to above “standard” case. Radial gradient of the plasma density profile was increased particularly in SOL, and both $n_{i}^{sep}$ and $n_{e}^{sep}$ were increased from $\rm 1.6x10^{19}$ and $\rm 2.1x10^{19} m^{-3}$ to $\rm 2.4x10^{19}$ and $\rm 2.9x10^{19} m^{-3}$, respectively. Since $n_{He}^{sep}$ $\rm \sim 1x10^{18} m^{-3}$ and $n_{Ar}^{sep}$ $\rm \sim 2x10^{17} m^{-3}$ near the separatrix, $n_{e}^{sep}$ was $\sim$25$\%$ larger than $n_{i}^{sep}$. Radial gradient of the temperature profile was increased near and inside the separatrixa. $c_{He}^{div}$ was increased to $\sim$15$\%$ and $\sim$10$\%$ in the inner and outer divertors, respectively. On the other hand, $c_{He}^{edge}$ = 7-9$\%$ was slightly increased. Since plasma performance such as $P_{fus}$ and $HH_{98y2}$ for the JA DEMO is based on system code results with $c_{He}$ = 7$\%$ in the main plasma, the plasma design is consistent with above simulation results, but it is necessary to avoid higher $c_{He}^{edge}$.
For the fuel particle exhaust, neutral and gas pressures ($P_{D0}$, $P_{D2}$) in the divertor were evaluated at exhaust slots of the dome and in the sub-divertor. For the “standard” case (without include neutral-neutral collisions, NNC), total neutral pressure ($P_{D}$ = $P_{D0}$ + $P_{D2}$) was increased from $\sim$2 to $\sim$3 Pa at the exhaust slots, and from $\sim$1 to $\sim$1.8 Pa in the sub-divertor, with increasing $\rm D_{2}$ puff rate from $\rm 4.8x10^{22}$ to $\rm 9.6x10^{22} D/s^{-1}$. Effects of NNC on the particle exhaust and detachment are shown.Speaker: Dr Nobuyuki Asakura (National Institutes for Quantum, Science and Technology (QST), Naka Institute) -
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Design of the divertor and power exhaust scenarios development for the Divertor Tokamak Test facility
The new high field superconducting divertor tokamak test facility (DTT) [1] presently under construction is devoted to specifically study power exhaust solutions in regimes as close as possible to those foreseen in DEMO fusion reactor in terms of power crossing the separatrix, $P_{sep}/R$, and heat flux decay length, $\lambda_q$. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. The divertor has being designed to test and compare a wide set of different power exhaust solutions, in particular the standard SND configuration and with high priority some of the most promising ADCs, like the X divertor (XD) and the “hybrid Super-X/long leg SN” but not excluding the possibility to test also the SnowFlake (SF) one. Additionally, the negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs and to easier power exhaust management.
Considering the wide requirements in term of divertor configurations acceptance the definition and optimization of the divertor shape has been done by an extensive power exhaust modelling with the 2D edge code SOLDGE2D because it allows to manage all above mentioned divertor configurations without any constraint in divertor (and wall) shape. Some specific analyses have been also done with the 2D edge code SOLPS-ITER code, like in studying the effect of the divertor dome in presence of drifts. Divertor optimization has been done at the maximum additional power presently foreseen for DTT (45 MW), the toroidal field (6 T) and plasma current achievable for the various configurations (5.5 MA in SND, 4.5 MA in XD) and a density corresponding to a Greenwald fraction of about 0.5 in the SND case. Transport profiles have been validated in present experiments and tuned to provide a $\lambda_q$ in agreement with available scaling laws, they have been kept constant in modelling all H-mode configurations. Considering the technological limits imposed on divertor shape by W monoblocks, it has been found that the best performances (lowest impurity content to achieve detachment by impurity seeding) between all configurations can be obtained with a wide divertor using the dome as third target and pumping slots for neutral compression.
Starting from the preliminary requirements, this presentation describes the compatibility of the various magnetic configurations with the technological constraints; the modelling activity to define the divertor shape and the different operating scenarios.References
[1] DTT - Divertor Test Tokamak - Interim design report, R. Martone, et al., (editors), ENEA - Frascati (2019). https://www.dtt-project.enea.it/downloads/DTT_IDR_2019_WEB.pdf.Speaker: Paolo Innocente (Consorzio RFX - CNR) -
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A tightly-baffled, long-legged divertor concept for DEMO and its potential test in TCV
Safe plasma exhaust with high core performance is one of the remaining challenges on the development path towards a fusion reactor. Alternatives to conventional single-null divertors are assessed as risk mitigation. While multiple null divertors face severe technological challenges, long-legged divertors emerge as a promising option [1] where combination with tight baffling may further increase the heat exhaust potential [2].
A proposed upgrade of the TCV tokamak aims at proof-of-principle experiments of such a tightly-baffled, long-legged divertor (TBLLD). In a first phase, new in-vessel components and diagnostic enhancements will provide a well-diagnosed, tightly baffled outer divertor leg to assess TBLLD’s ability to increase the plasma neutral interaction and impurity radiation and, thereby, facilitate detached divertor operation. Experiments will also investigate the stability margin between divertor re-attachment and over-mitigation together with passive mechanisms and active technologies to operate within those margins. Initial SOLPS-ITER simulations predict that the first phase should demonstrate a significant increase in the power handling capability. Relying upon intrinsic carbon radiation alone or employing nitrogen seeding should both lead to at least a 5-fold increase of the power handling capability over the unbaffled TCV divertor. Experiments in the first phase will test the divertor models that are used to prepare the second upgrade phase and assess the plasma exhaust potential of the TBLLD concept in DEMO. In this second upgrade phase, the TBLLD concept will be applied to the entire TCV divertor in combination with a promising high-performance plasma core with up-down symmetric configurations and negative triangularity shaping being currently considered as options.
Initial scoping studies investigate the application of such a TBLLD concept into the EU DEMO design without significant changes to the conventional single-null equilibrium solution. This is achieved by using the full depth of the space defined by the tritium breeding blanket, similar to the proposed ARC fusion pilot plant design, but without the challenging, and costly, complications of that magnetic topology [3].
[1] F. Militello, et al., Nucl. Mater Energy 26 (2021) 100908.
[2] M. Umansky, et al., Nucl Fusion 60 (2020) 016004.
[3] A.Q. Kuang, et al., Fusion Eng. Des. 137 (2018) 221.Speaker: Holger Reimerdes (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland)
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12:50
Lunch Board Room A
Board Room A
IAEA Headquarters
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Poster Session I (DEMOs & Next Step Facilities) Board Room A
Board Room A
IAEA Headquarters
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Impurity parallel force balance in the SOL using the Zhdanov closure in Soledge3X-EIRENE simulations of WEST discharges
Next steps machines such as ITER and DEMO will face unprecedented challenges related to heat exhaust. Impurity seeding will play a key role in spreading power on a sufficiently large surface area. Simulating impurity radiation in a given scenario requires i) radiation functions for the different ion stages and ii) the spatial distribution of the ion densities. The latter results from the location of the injection and from a competition between parallel and perpendicular transport. Correct interpretation of current experiments and extrapolation to larger machines all the more so with advanced divertor configurations, have to rely on a solid theoretical basis for impurity dynamics. This work aims at strengthening this basis by evaluating the sensitivity of parallel impurity force balance, which plays a key role on setting the poloidal impurity distribution as well as divertor leakage, to the fluid model used for impurities.
Multifluid equations have been derived by Zhdanov in the 1980’s [1], and have been recently revisited by several groups [2,3]. These equations are valid beyond the trace approximation and provide expressions for the friction force, thermal forces and energy exchanges between species from kinetic theory. This closure is implemented in the SolEdge3X-EIRENE code package [4]. In this contribution, WEST 2D transport simulations performed in support of spectroscopic measurements [5] are shown to exhibit strong in/out asymmetry for high Z oxygen ions (oxygen is used as a representative impurity throughout this work), which complex spatial patterns. The parallel force balance is analysed to explain these findings, partly related to the Z dependences of forces, and evaluate the sensitivity of the parallel balance to the collisional closure. The importance of non-trace effects is evaluated in these conditions. In contrast to most of other edge code packages, SolEdge3X solves one energy equation per species and the relative temperature deviations between species are found to reach ~ 20% in the WEST simulations, with the sheath model used. The impact of these deviations on the parallel force balance is evaluated using recent extensions of the closure [2], properly accounting for the temperature difference between species in the calculation of parallel forces.
[1] V. M. Zhdanov, Transport processes in multicomponent plasma, Taylor and Francis, London, 2002.
[2] M.Raghunathan et al., 2021 Plasma Phys. Control. Fusion 63 064005;
[3] S. O. Makarov et al., 2021 Physics of Plasmas 28 062308.
[4] H. Bufferand et al., 2021, Nucl. Fusion 61 116052.
[5] C.C. Klepper et al., submitted to Plasma Phys. Control. Fusion.Speaker: Yannick Marandet (PIIM, CNRS/Aix-Marseille Univ., Marseille, France, EU) -
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Simulation study of the density threshold for the reverse of the detachment priority between inner and outer divertor for quasi-snowflake configuration
In future fusion reactor, it is critical to achieve detachment for the divertor, in order to protect the targets and sustain the steady state operation. By introducing an additional X-point in the poloidal field, the snowflake divertor [1] has larger flux expansion and connection length compared with the single-null divertor. Thus, the snowflake divertor is expected to promote the detachment, and considered as a promising candidate for the future fusion reactor. In this work, the quasi-snowflake divertor is studied. In the quasi-snowflake divertor, the additional X-point is at a distance from the main X-point and thus only two strike points exist on the two divertor targets. Although the increase of flux expansion and connection length is not as large as that for exact/near-exact snowflake divertor, the requirement of the ability of poloidal field coils and the complexity in the engineering can be reduced. During the simulation study for China Fusion Engineering Test Reactor (CFETR), in addition to the promotion of detachment [2], it is found that the outer divertor could be detached in prior to the inner target under relatively low upstream density [3].
The reverse of the detachment priority between inner and outer target is further studied [4]. A series of quasi-snowflake divertor configurations with similar flux expansion in inner divertor and increasing flux expansion in outer divertor, are created using EFIT. By the scan of upstream density in the SOLPS-ITER simulation, the density threshold is identified for each quasi-snowflake divertor, below which the outer divertor is detached in prior. The density threshold is found positively related to the flux expansion in outer divertor. The existence of the density threshold is due to the competition between the neutral compression, which benefits the detachment in the inner divertor, and the flux expansion, which benefits the detachment in the outer divertor. At the low upstream density, the magnetic flux expansion plays a dominant role, while at high density, the neutral density gradually manifests and finally overcomes the former. The seeded neon impurities is more likely to radiate power at the side which is detached in prior, thus impurity seeding acts as an amplifier of the asymmetry between the inner and outer divertor.
To explore the influence of the machine size on the detachment priority between inner and outer divertor, SOLPS-ITER simulations are carried out further for the artifact EAST quasi-snowflake configurations. The influence due to E×B drifts on the detachment asymmetry is also studied. The details will be reported in the conference.Reference
[1] D.D. Ryutov, Phys. Plasma 14 (2007) 064502.
[2] S.F Mao et al., J. Nucl. Mater. 463 (2015) 1233.
[3] M.Y. Ye et al., Nucl. Fusion 59 (2019) 096049.
[4] X.L. Ruan, et al., Simulation study of the influence of flux expansion on the detachment sequence of HFS and LFS divertor targets (in preparation).Speaker: Xinglei RUAN (USTC) -
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Measurements of impurity flows and line-radiation in the W7-X scrape-off layer
Characterizing the scrape-off layer (SOL) transport of Wendelstein 7-X (W7-X) is essential to assess the efficiency of its unique exhaust concept, the island divertor configuration. Insights into the SOL dynamics in attached and detached conditions can be gained by the measurement of particle flows. Investigations of impurity flow velocities and line-radiation have been carried out with the Coherence Imaging Spectroscopy (CIS) diagnostic, featuring 2D spatially resolved measurements that led to the first detection of the 3D counter-streaming flow pattern of the W7-X SOL [1]. The impurity monitored by CIS is C2+, selected via the C-III transition line at 465 nm (2s3p3P°→2s3s3S). Its line-emission intensity, integrated over the entire camera view, is observed to be linearly proportional to the total plasma radiated power (Prad) in both attached and detached plasmas for non-seeded conditions. This linear relationship, together with a multi-machine scaling for Prad [2], is exploited to link the C-III intensity to the line-averaged density, demonstrating experimentally that the plasma density can be used as an actuator for the carbon line-radiation. The related C2+ velocity exhibits a strong dependence on the SOL density, while the SOL input power has no direct influence on the velocity magnitude. In attached plasmas, both the velocity and the SOL density increase with increasing line-averaged density. The tendency reverses during detachment, in which both quantities decrease by at least a factor of 2 [3,4]. The sharp drop in velocity, together with a rise in line-emission intensity, is reliably correlated to the achievement of the detached state and can be used as one of its signatures [3]. Another distinctive feature of the measured C-III radiation during the transition to detachment is the appearance of localized emission areas around the X-point regions of the island chain, as predicted by EMC3-Eirene [5]. In the measurement domain, the impurity flow velocity appears to be well-coupled with the main ion one, thus implying the dominant role of impurity-main ion friction in the parallel impurity transport dynamics [6]. Using the C2+ impurity flow as a proxy for the bulk plasma one, the CIS results are interpreted with the help of EMC3-Eirene simulations, but their major trends are already explainable with a simple 1D fluid model, based on the continuity equation and the SOL power balance. At the same time, EMC3-Eirene modelling is not able to entirely capture the measured tendencies in the CIS flow velocities, due to the incompleteness of the physical model currently used in the code. A key missing physics aspect is the E×B drift, which is shown to substantially influence the measured flows, especially in the location of their stagnation regions [7].
References
[1] V. Perseo et al., Nuclear Fusion, vol.59, 2019.
[2] G.F. Matthews et al., Journal of Nuclear Materials, 241-243, 1997.
[3] V. Perseo et al., Nuclear Fusion, vol.61, 2021.
[4] F. Reimold et al., 28th IAEA FEC 2020, IAEA-CN-286-1296, 2021.
[5] Y. Feng et al., Nuclear Fusion, vol.61, 2021.
[6] Y. Feng et al., Nuclear Fusion, vol.49, 2009.
[7] D.M. Kriete et al., in preparation, 2022.Speaker: Dr Valeria Perseo (Max-Planck-Institute für Plasmaphysik) -
14
Assessing Alternative Divertor Configurations for Power Exhaust in EU-DEMO
The realization of EU-DEMO will pose unprecedented challenges to the power exhaust system. Its plasma-facing components will have to cope with a steady-state load of about 450 MW (fusion + additional heating). It is expected that creating and controlling a proper radiating edge region will isotropically dissipate ~ 300 MW, with the remaining 150 MW flowing in the Scrape-off Layer towards the targets. The increase in machine size with respect to ITER (RDEMO ~ 9m, RITER ~ 6 m) compensates only partially for the larger power to exhaust, resulting in an environment more severe than the one expected for ITER. On the top of that, requirements to keep a sufficient Tritium Breeding Ration (~1.2) pose severe constraints on the First Wall design.
There are suggestions that operating a divertor extrapolated from the ITER design could be possible, at least in steady state. However, such predictions are not conclusive. They are sensitive to the Scrape-off Layer power decay length, which is itself a largely uncertain quantity (ranging between <1 and >10 mm, according to current models). As a backup option, EUROfusion is exploring several Alternative Divertor Configurations (ADC) including X Divertor, Super-X, Double Null and Snow-Flake configurations. On the one hand, preliminary studies showed that many ADCs have a larger acceptable operating space than the standard SN, potentially allowing for a more robust and easily controlled reactor. On the other hand, such configurations often present challenging engineering constraints, and may become worth pursuing only if sufficient advantages over standard SN can be reliably demonstrated. Such task requires considerable modeling effort.
Here we present the status of computational studies on ADCs in Europe. Much work was performed using the fluid neutral option available in the SOLPS-ITER code. This allows capturing most of the magnetic equilibrium-dependent effects (e.g., flux flaring, connection length and grazing angle). First results suggest that detachment can indeed be obtained more easily. Comparison with simple analytical models (e.g., by Lengyel) shows that radial transport plays a significant role in reducing the impurity concentration needed to obtain a sufficient radiation level, by increasing the available radiating volume. This contributes to favor configurations with large connection length. Recently, studies with kinetic modeling of neutrals were also started. Although more computationally demanding, they allow increasing the accuracy of the physical description, by (i) including a proper treatment for hydrogenic molecules, (ii) allowing for a more detailed description of the chamber wall and reflection processes, and (iii) de-coupling the ion and neutral temperatures. We also discuss the status of such modeling and highlight planned studies. Finally, recent developments in the available computational tools, including advanced fluid neutral models and extension of the fluid mesh up to the vessel wall, show potential for strong increase in the accuracy of edge plasma studies. We also discuss the development level and first applications of such advances, focusing on consequences for ADC studies.Speaker: Fabio Subba (Politecnico di Torino) -
15
Developing tools to accelerate divertor design and full-power operation
We aim to address the design and operation challenges of future tokamak fusion reactors by developing tools which combine data and simulations with automation and uncertainty quantification techniques. We will present recent progress towards this goal in model development, workflow automation, and plans to develop and validate these tools in collaboration with current and near-term experimental facilities including the DIII-D national fusion facility.
The divertor is a critical part of any high-power tokamak device and an integral part of the overall design, strongly impacting the plasma operating space and choice of poloidal field coil configuration. The time required to identify and refine performant divertor designs has the potential to delay the whole design process. Once a tokamak is constructed, ramp-up to full power operation is dependent on managing risks of damage to the divertor and other plasma facing components. As with design, the time required to assess the safe operating space of the divertor with confidence could delay the start of full power operation. We aim to accelerate these timelines by automating the assimilation of data and simulations and their uncertainties.
The physics basis of our tools is provided by two open-source codes: The well-established UEDGE (https://github.com/LLNL/UEDGE) and a new code, Hermes-3 (https://github.com/bendudson/hermes-3/). The Hermes-3 code has been developed based on BOUT++ and is capable of multi-fluid transport and turbulence calculations in tokamak single and double-null geometries. We will present these tools, their advantages and disadvantages, and their potential role as part of a larger multi-fidelity suite of modelling tools to accelerate the design and operation of tokamak fusion reactors.Prepared by LLNL under Contract DE-AC52-07NA27344.
Speaker: Dr Tom Rognlien (LLNL) -
16
Assessing simple models for density build up and impurity exhaust in the island divertor of W7-X
In the previous operation campaigns of W7-X stable, detached divertor plasmas were demonstrated [1,2,3,4]. These plasma also showed particle exhaust that enables high density steady-state operation and provided good impurity retention [5]. Hence, the initial results from W7-X operation show a potential of the island divertor concept used in W7-X for a reactor. Spectroscopic measurements and other edge diagnostics indicate the existence of a high-density divertor regime in W7-X ($n_{div} > n_{up}$). Such a regime was not observed in the predecessor W7-AS, but its occurrence had been predicted for W7-X by modeling [6,7]. However, the achieved neutral divertor pressures so far have been limited to low values of $< 0.1 Pa$. Given the crucial role of the neutral pressure as a design variable of ITER divertor [8] and its exhaust regime, this makes the understanding of the density build-up in the island divertor a crucial task, in particular to assess the particle exhaust properties and their scaling towards a reactor device. It is important to understand what sets and limits the neutral pressures and to identify the key physics parameters that drive the density evolution. To this end simplified models and comparison to modeling is used to assess their applicability to the W7-X data and their usefulness to understand the relevant dynamics in the density-build up in the island divertor. Adaptations of the two-point model accounting for cross-field transport channels in the island divertor [2] will be used to compare with experimental measurements of the density build-up in the divertor in density ramp experiments. The model implies a crucial role of the pressure loss factors and its dependence on temperature as well as on the fieldline pitch $\Theta$ of the island. Both quantities are expected to be fundamentally different in the island divertor compared to a tokamak.
The relevance of the divertor density with respect to the impurity exhaust, most notably for He, will be discussed in light of the recent analysis of the impurity retention capabilities of the island divertor [9] and the more challenging core transport properties of stellarators due to unfavorable neoclassical transport.
[1] M. Jakubowski, Nuc. Fus. 61 (2021)
[2] O. Schmitz, Nucl. Fus. 61 (2020)
[3] D. Zhang, Phys. Rev. Let. 123 (2019)
[4] F. Effenberg, Nucl. Fus. 59 (2019)
[5] V. Winters, PSI 2020
[6] Y. Feng, Nucl. Fus. 46 (2006)
[7] Y. Feng, Plasma Phys. Control. Fusion 53 (2011)
[8] A. Kukushkin, Fus. Eng. and Design 86 (2011)
[9] V. Winters, Nucl. Fus. (submitted)Speaker: Felix Reimold -
17
Implementations of parallel ion viscosity in SOLEDGE3X and their impact on edge plasma turbulence
Despite several theoretical approaches and few pioneering modelling results [1], the onset of the H-mode transport barrier around the separatrix remains a rather open issue. Understanding precisely the physics behind this enhancement on the confinement and being able to reproduce it in numerical simulations is still a major objective to analyse the performances of nowadays tokamaks and prepare the operation of future devices such as ITER. In that perspective, a dedicative effort has been made over the last decade to develop the 3D turbulent code SOLEDGE3X-EIRENE to tackle the challenge of L-H transition modelling.
Plasma rotation and poloidal velocity shear is a potential player in turbulence reduction implied in L-H transition. Hence, a special care must be taken to investigate the various mechanisms controlling plasma rotation. In this contribution, we will discuss the impact of parallel ion viscosity through its impact on momentum balance. In the fluid description, the parallel ion viscosity terms are the ones that drive the rotation towards the neo-classical value. The reference Braginskii formula is known to lack features to reproduce properly the neo-classical regime even in the high collisional Pfirsch-Schlüter regime [2]. Therefore, several corrections have been proposed in the literature to improve it in order to recover first the Pfirsch-Schluter limit [3] and sometimes even lower collisionality regimes [4]. SOLEDGE3X implements several of these expressions to test their ability to recover numerically neo-classical rotation. We report on their numerical implementation and stability. The effect of the rotation on turbulence in the simulations is also discussed.[1] S. Ku et al., Phys. Plasmas 25, 056107 (2018)
[2] P. Helander & D. Sigmar, Collisional Transport in Magnetized Plasmas, Cambridge University Press (2005)
[3] A. Mikhailovskii, V. Tsypin, Beitr. Plasmaphys. 24, 335-354 (1984)
[4] T. Gianakon, S. Kruger, C. Hegna, Phys. Plasmas 9, 536 (2002)Speaker: Hugo Bufferand (CEA) -
18
Kinetic trajectory simulation method for plasma-wall interaction of multi-component, electronegative, and dusty plasmas
The kinetic simulation method has been employed to study the plasma-wall interaction mechanism in various plasma conditions: multi-component plasmas, electronegative plasmas, and dusty plasmas. The negative species are considered to enter the sheath region from presheath side with truncated Maxwellian distribution, and the ions satisfy the Bohm and Bohm-Chodura conditions. The sheath conditions have been extended for the cases considered and various plasma-wall transition properties (space charge density, sheath potential, phase-space trajectory, and particle flux towards the wall) are found to be affected. The simulation model is applicable to various situations of interest in divertor region as well as in bounded plasmas. The presence of magnetic field and electrode biasing significantly alters the scale length of the Debye sheath region and Chodura layer as well. The dust charging, magnitude of ion drag force, and levitation of charged dust grains in the transition region depend on the biasing voltage and size of the grains. The dust particle acquires a negative charge at the particle injection boundary and becomes positively charged close to the electrode due to electron depletion in the Debye sheath region. The stable levitation distance, dust charge, and required electric field at that point have been estimated for the varying negative voltage applied to the electrode.
Speaker: Raju Khanal (Tribhuvan University) -
19
Modeling the effect of drifts on HL-2M snowflake in-out divertor heat loading by SOLPS-ITER
Divertor detachment operation is of critical importance for future long pulse high power tokomaks, such as ITER and CFETR[1]. Partially detached divertor conditions are foreseen for ITER and DEMO. In conventional standard divertor (SD) operation, the effect of drifts on SD detachment is very strong, and the outer target heat loading is very higher than inner target with favorable-Bt discharge[2]. The root reason is the polodial drift flux driving a lot of recycling plasma and radiated impurity into inner target region along private flux region. Currently, the experiment results have demonstrated that the drifts also have a great influence on the snowflake divertor detachment in TCV[3]. But, its detail physical mechanism is awaiting a better theoretical understanding of the relevant physical picture.
HL-2M allow operation with various advanced snowflake divertor configurations such as SF+ and SF-, and provides a good platform for studying advanced snowflake divertor plasma physics[4]. HL-2M snowflake divrtor was designed by using SOLPS5.0[4]. The SOLPS5.0 modeling results without drifts showed that the inner target heat flux (qprep~ 15MWm-2 )and Te (Te ~60eV )are very higher than them in outer target plate (qperp~1MWm-2, Te~2eV)[5]. Although the outer snowflake divertor with large magnetic expansion can effectively reduce the outer target heat loading, the inner divertor target heat loading is very high. In fact, the drifts may has a very stronger effect on snowflake diervtror detachment/in-out asymmetry than SD. Therefore, in this work, we will investigate the influence of drifts on HL-2M snowflake diervtor in-out target plate heat loading by employing the latest edge plasma code SOLPS-ITER[6] considering 100% all drifts and current.
References
[1] A.S. Kukushkin, et al. Consequences of a reduction of the upstream power SOL width in ITER, J. Nucl. Mater. 438 (2013) 6–10.
[2] V. Rozhansky, et al. Contribution of E × B drifts and parallel currents to divertor asymmetries, Nucl. Fusion. 52 (2012) 103017.
[3] B.P. Duval, et al. Enhanced EXB drift effects in the TCV snowflake divertor, Nucl. Fusion. 55 (2015) 123023.
[4] G.Y. Zheng, et al. Investigations on the heat flux and impurity for the HL-2M divertor, Nucl. Fusion. 56 (2016) 126013.
[5] H. Du, et al. Exploring SF- in-out asymmetry and detachment bifurcation in HL-2M with E × B by SOLPS, Nucl. Mater. Energy. 22 (2020) 100719.
[6] X. Bonnin, et al. ITER divertor plasma response to time-dependent impurity injection, Nucl. Mater. Energy. 0 (2016) 1–6.
Acknowledgments
This work was supported by National Natural Science Foundation of China under Grant Nos. 12175054; National Key R&D Program of China Nos. 2022YFE03180300, 2018YFE0301101.Speaker: Hailong Du (Southwestern Institute of Physics) -
20
Simulation study of the E×B drift effect on the transport of plasma and sputtered impurities in scrape-off layer
It is essential to control the plasma flow onto the divertor targets to avoid unacceptable heat load and erosion, and the contamination due to the sputtered impurities to avoid the degradation of the performance of confined plasma. Thus, comprehensive understanding of the plasma and impurity transport in the scrape-off layer (SOL) is necessary for exploring the divertor solution for the future reactor. Recently, it has been demonstrated that the E×B drift can affect the plasma [1] and impurity [2] transport. For the purpose to improve the understanding for the effect of E×B drift effect on the SOL plasma transport as well as the sputtered impurities from the divertor targets, simulation studies [3,4] have been carried out using SOLPS-ITER [5] and DIVIMP [6] based on the EAST configuration.
For the main ion species, SOLPS-ITER simulations under favorable and unfavorable toroidal field (BT) are performed by scan of the PSOL and upstream density. Two boundary lines have been identified on the PSOL-ne,sep plane. Between the boundary lines, the double-peaked characteristic can be observed in the density profile at inner target for favorable BT and outer target for unfavorable BT. The appearance of the upper boundary is qualitatively consistent with the EAST experiment [7], and some evidence for the existence of the lower boundary can also be found in the experimental data. Regarding the appearance of the second peak, the synergetic effect of poloidal and radial E×B drifts is considered as the main mechanism. The radial E×B drift results in the effective particle sink in near-SOL region and the source in far-SOL region, and the effect is enhanced due to the obstruction to the plasma flow by the poloidal E×B drift in the near-SOL region. Furthermore, the simulations with neon seeding also performed. It is found that the impurity seeding affects the target density profile on the contrary way to PSOL.
Within the simulating background plasma with neon seeding, the transport of the sputtered tungsten impurities from the divertor targets is studied by DIVIMP. The E×B drift is included in the simulation by introducing the calculated drift velocities of the tungsten impurities. The E×B drift is found to enhance the core contamination of the tungsten impurity, and the effect is more significant under unfavorable BT. Detailed analysis of the effect of E×B drift on the transport of tungsten impurity will be presented in the conference.Reference
[1] H.Q. Wang, et al. Phys. Rev. Lett. 124 (2020) 195002.
[2] I.Y. Senichenkov, et al. Plasma Phys. Control. Fusion 61 (2019) 045013.
[3] J. Guo, et al. Simulation study of the influence of upstream density and power into scrape-off layer on the double-peaked density profile at divertor target. (Submitted to Nucl. Fusion)
[4] J. Guo, et al. Simulation study of the effects of E×B drift on core tungsten concentration. (in preparation)
[5] X. Bonnin, et al. Plasma Fus. Res. 11 (2016) 1403102
[6] P.C. Stangeby and J.D. Elder. Nucl. Fusion 35 (1995) 1391.
[7] J. Xu, et al. Nucl. Fusion 61 (2021) 096004.Speaker: Shifeng MAO (University of Science and Technology of China) -
21
SPARC Diagnostics for Use in Plasma Control and Divertor Physics Studies
The SPARC tokamak is a compact, high-field short pulse device (B0 = 12.2 T, R0 = 1.85 m, τflattop = 10 s) that plans to begin operations in mid-2025. It will execute a series of mission-driven campaigns to close science gaps to inform the design of the ARC fusion pilot plant to begin operation in the early 2030’s. Accomplishing this requires a versatile plasma diagnostic set for use in real-time control and inter-shot learning. Diagnostic requirements are driven in part by SPARC’s Advanced Divertor Mission, which seeks to answer seven key questions on divertor physics and power exhaust at high, ~10 GW/m2, parallel SOL heat fluxes and edge plasma conditions matching those required for a pilot-plant. Practical constraints limit boundary diagnostic performance, including the timely integration of embedded sensors in a high-temperature, Tlimit ~ 400 degC, high-neutron flux (Pneut/Spl ~ 2 MW/m2), tritiated environment and observing a closed divertor using low-field side, port plug-based measurements.
The full SPARC diagnostic set consists of over 40 sub-systems, multiple of which are relevant for edge and divertor physics. These include magnetics used to reconstruct a range of divertor geometries, main-chamber and divertor neutral pressure measurements, wide spectral range spectroscopy from VUV to NIR to track impurities and characterize hydrogen isotopes, visible and IR camera imaging, bolometry, Langmuir probes, reflectometry for edge density profiles and temperature and strain sensing embedded in PFCs. This contribution focuses on the details of the diagnostics while a companion contribution, “Utilization of SPARC to investigate divertor solutions for fusion pilot-plants” A.Q. Kuang, focuses on SPARC divertor design.
While details of SPARC’s diagnostic are subject to evolution, the type and range of anticipated measurements is known. Day-to-day operation will require monitoring and controlling heat flux to plasma facing components. This will use several hundred embedded temperature measurements, mostly via stainless-steel sheathed, Type-N thermocouples, complemented by NIR/IR camera imaging from up to ten toroidally and poloidally spaced imaging port-plugs. Heat exhaust will be mitigated through injection of fuel and impurities, monitored by a 200+ channels of bolometry, micro-Penning and crystal cathode gauges and visible and VUV spectroscopy measurements in the inner and outer divertor. Flux loop and poloidal field sensor probes will reconstruct the x-point and strike point locations, with strike point sweeping employed to spread heat over 10’s of cm of poloidal PFC surface, per divertor. Interpretation of these measurements will be facilitated through modeling using the Heat Flux Engineering Analysis (HEAT) toolkit.
Dedicated divertor physics studies, such as determining access to and control of detachment, requires many of the tools used in heat flux control, but pushes requirements on channel count, time resolution and flexibility in observing a range of divertor geometries and conditions. Swept voltage Langmuir probes are not envisioned for high heat flux regions of the divertor, but will be used for novel x-point target geometries that use separate PFC surfaces for strike points. Requirements for visible impurity spectroscopy to enable estimates of tungsten erosion and helium enrichment are also uniquely driven by the Advanced Divertor Mission.Speaker: Matthew Reinke (Commonwealth Fusion Systems)
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16:30
Break Board Room A
Board Room A
IAEA Headquarters
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Discussion Session: DEMO & Next Step Board Room A
Board Room A
IAEA Headquarters
Conveners: James Harrison (United Kingdom Atomic Energy Authority), Nobuyuki Asakura (National Institutes for Quantum and Radiological Science and Technology (QST))
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SOL Board Room A
Board Room A
IAEA Headquarters
Convener: Anthony Leonard (USA)- 22
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Plasma-neutral interaction processes in ITER and medium-sized tokamaks from SOLEDGE3X full-vessel boundary plasma simulations
Interactions between plasma and neutrals in the volume and in the vicinity of the divertor and vessel wall surfaces largely determine particle and heat exhaust capabilities of plasma scenarios, especially for reactor-relevant detached regimes. Identifying and understanding the key atomic and molecular processes involved in these interactions is critical to modelling efforts.
In this work, we propose an assessment of the mechanisms at play in a wide range of edge plasma conditions. Discharges for ITER and two current medium-sized tokamaks (JET and AUG) are simulated with the SOLEDGE3X [1] code in 2D transport mode, coupled to the EIRENE kinetic neutrals code, including recently developed detailed output diagnostics for decomposing and studying the contributions from each process included in the plasma-neutral interaction model. Of particular interest here are ion-molecule collisions (elastic and charge-exchange), Molecule Assisted Recombination (MAR), and neutral-neutral collisions. Since the SOLEDGE3X code uses a grid covering the entire vacuum vessel volume with a realistic wall geometry, the produced plasma solutions include the divertor as well as the far Scrape-Off-Layer (SOL) up to the first wall, providing the plasma conditions and fluxes at both locations.
First, a throughput scan is performed for ITER, considering Pre-Fusion Operation Phase 1 (PFPO-1) scenarios with 20 MW of SOL power input. A notable result from this scan is that the molecule charge exchange, while almost negligible in attached cases, plays a key role when partial detachment is reached.
Second, to assess the impact on main chamber recycling of the flattening of the far-SOL density profile observed in current machines under high density detached divertor conditions [2], particle and heat diffusivity coefficients are increased in the far-SOL for 20MW and 60MW cases, mimicking turbulent filamentary transport. Increases in temperatures, particle and heat loads on the first wall of the main chamber are found (e.g. by a factor of up to 4 for the integrated heat load) from additional atom charge exchanges with ions at higher energies. For the cases with the highest far-SOL diffusivity assumptions, even the divertor solution starts being affected and fluxes reduced (e.g. by up to -35% for the total perpendicular heat flux density).
Third, investigation of the differences in the atomic and molecular processes involved in high-throughput simulations on machines of similar geometries but different size is presented, using the AUG and JET devices as examples. This exercise highlights specificities of ITER versus smaller devices, in part due to a change in the ratio of the contribution of the molecular ion H2+ dissociation to the atom ionisation in the particle and electron energy sources near the targets.[1] H. Bufferand et al., Nuclear Fusion 61 (2021) 116052
[2] N. Vianello et al., Nucl. Fusion 60 (2020) 016001Speaker: Nicolas RIVALS (CEA) -
24
Modeling transient edge plasma processes with dynamic wall recycling
A self-consistent 2D model is presented for transport in boundary plasma and plasma-facing material walls. Plasma dynamics in the domain is represented by a 2D collisional plasma fluid model in the edge-plasma code UEDGE [1], and transport of hydrogen and heat in the wall material is represented by a system of 1D (into the wall) reaction-diffusion equations solved in the code FACE [2]. To account for variation of parameters along the wall, in the coupled model multiple instances of the FACE code run in parallel. The coupled model provides a tool for investigating a range of dynamic plasma-material interactions phenomena in 2D. For studies of edge plasma with active wall, two applications are considered here: (i) strike-point sweeping, and (ii) ELM-pulse simulation.For strike-point sweeping, coupled calculations are applied to investigation of the impact of heat and hydrogen transport in the material wall on the divertor plasma and target heat load during sweeping of the target strike-point for parameters of high-power tokamak operation [3]. The modeling shows that for realistic sweep parameters frequency 10 Hz and amplitude 10 cm, the temperature on the plate can be maintained below 1500 K. However, sputtering of a target plate when sweeping without plasma detachment remains an unresolved problem under the conditions investigated. For ELM simulations, release of hydrogen from the wall can trigger transition to detachment in divertor [4]. During an ELM strike when the divertor plasma is attached, hydrogen is released from the target plate near the strike point and absorbed in the far SOL wall regions. On the other hand, for semi-detached divertor plasma, hydrogen is absorbed by the divertor plate during the ELM strike and released afterwards. During an ELM strike, hydrogen retention is affected up to the depth ~1 μm, while the plate is heated through ~100 μm. Overall, self-consistent simulations of edge plasma with an active wall add a new important modeling capability for systems with long pulses such as a tokamak-based fusion reactor.
[1] T.D. Rognlien et al., J. Nuc. Mat. 196, 347–123 (1992); [2] R.D. Smirnov et al., Fusion Sci. Technol. 71 (2017) 75; [3] M.V. Umansky et al., Contr. Plasma Physics (2022) doi: 10.1002/ctpp.202100156; [4] R.D. Smirnov et al., Phys. Plasmas 27 (2020) 032503.
*Work supported by US DoE contract DE-AC52-07NA27344 at LLNL and award DE- SC0018302 at UCSD as part of SciDAC-PSI.
Speaker: Dr Maxim Umansky (Lawrence Livermore National Lab) -
25
Experimental and numerical progress in the assessment of alternative divertor configurations in TCV and extrapolations towards higher power conditions
We will present recent progress of alternative divertor geometry studies in TCV, focusing primarily on the effects of total flux expansion (Super-X), outer target poloidal flux expansion (X-Divertor) and additional X-points near the target (X-Point Target) or near the primary X-point (Snowflake-Minus-LFS). These geometrical features are combined with improved divertor closure using the new TCV baffles and studies are directed towards operation at the highest available power levels.
The first baffled Snowflake L-mode experiments [1] show significant reductions in peak target heat fluxes due to both geometry and baffles. With nitrogen seeding, the X-point radiator present in comparable Single-Null discharges can be displaced outside the confined plasma, sitting stably between the primary and secondary X-point. This does not, however, provide benefits in terms of core confinement or core impurity screening, a behavior which is currently being interpreted using EMC3-EIRENE simulations and extended to higher power/SOL opacity conditions in both simulations and experiments. The beneficial effects of total flux expansion in Super-X L-mode plasmas fall short of Two-Point Model predictions, yet less so in the presence of baffles. These results are consistent with SOLPS-ITER simulations in an idealized setup [2] and more recent SOLEDGE2D-EIRENE runs closely matching the experimental setup [3], both highlighting the need for high-levels of divertor neutral trapping for optimized Super-X operation. Most high-power, H-mode detachment studies on TCV have focused on NBI heated X-Divertor and X-Point-Target divertors, both compared to a more conventional Single-Null. Stable, inter-ELM detachment is achieved with nitrogen seeding, with H98≳1 throughout and little difference in core properties as divertor geometry is varied [4]. Inter-ELM, outer target heat fluxes are reduced by factors ≳2 in the X-Divertor and X-Point-Target and, when combined with baffles and seeding, are shown to drop by ~95% as compared to the unbaffled, unseeded Single-Null. These experiments are currently being extended to small-ELM regimes and discharges combining the maximum level of NBI and ECRH heating on TCV and first results will be presented.
In addition to the use of transport codes to interpret and extrapolate these experimental findings, we will also discuss recent progress on TCV in the validation of first-principles turbulence simulations, insights gained on divertor turbulence and transport, and first studies on their dependence on divertor geometry.
[1] S. Gorno et al., “Power exhaust of the baffled SF-LFS divertor in TCV”, in preparation
[2] A. Fil et al., Plasma Phys. Control. Fusion 62 035008 (2020)
[3] C. Meineri et al., “Study of fully baffled super X L-mode discharges with PFR D2 fuelling on TCV”, in preparation
[4] H. Raj et al., “Improved Heat and Particle Flux Mitigation in High Core Confinement, Baffled, Alternate Divertor Configurations in the TCV tokamak”, in preparation
[5] D. S. Oliveira and T. Body et al., “Validation of edge turbulence codes against the TCV-X21 diverted L-mode reference case”, in pressSpeaker: Christian Theiler (EPFL-SPC) -
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Numerical Predictions of Divertor Power Sharing in Conceptual High Power Double Null Tokamaks
SOLPS-ITER is used to investigate power sharing between the upper and lower outer divertors in a balanced double null magnetic configuration, revealing that asymmetric divertor conditions drive SOL power flows which can result in a large power sharing imbalance between divertors. Double null (DN) or near double null magnetic geometries are potential configurations for a future tokamak reactor due to advantages such as more efficient fueling, favorable energy confinement, and shared power loads in two outer divertors which may justify DN configuration despite the reduction of available core plasma volume. Simulations of a compact high field tokamak (B_T = 6 T, R = 1.5 m, P_SOL = 4 to 50 MW) show that when one outer divertor is attached and the other is detached, power flows dominantly toward the attached divertor regardless of ion grad B direction. However, when both outer divertors are attached or both detached, power flows toward the divertor in the ion grad B direction.
These results challenge the DN concept of designing one outer divertor for dissipation and the other divertor for density control using pumping, because the resulting asymmetric divertor conditions create an unfavorable power sharing imbalance that favors the less dissipative pumped divertor. In order to achieve a desired level of power sharing, magnetic balance control will be critical in order to compensate for symmetry-breaking effects such as divertor shape asymmetries, pumping, and drifts.
Speaker: Dr Jonathan Yu (General Atomics)
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10:35
Break Board Room A
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SOL Board Room A
Board Room A
IAEA Headquarters
Convener: Anthony Leonard (USA)-
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Complete H fuel cycle with the island divertor in Wendelstein 7-X
A comprehensive analysis of the H fuel cycle, using experimental measurements aided by input from modeling, is presented for attached and detached plasmas. This analysis focuses on the particle transport processes in the Wendelstein 7-X island divertor is presented. This analysis allows an assessment of the status-quo and will quantify the optimization potential in particle collection, removal, and plugging for a future divertor design.
Particle control is of significant relevance for a nuclear fusion reactor as it affects a wide range of operational and safety aspects. The hydrogen fuel cycle inside the plasma vessel affects core plasma confinement, in-vessel components lifetime, fuel and therefore also tritium retention. Fueling and exhaust provide a path to high-density regimes, which are necessary for high performance and detached plasmas. The mission of the optimized stellarator Wendelstein 7-X is to demonstrate steady state, high performance plasmas. First experiments with the island divertor, a novel concept for heat and particle exhaust deliver exciting results including complete, stable detachment 1,2.
The particle confinement time is relatively short, τP = 0.26 s 3. As particles leave the plasma, 99 % of the outflowing ion flux is neutralized at the divertor target plates, according to EMC3-EIRENE modeling4. These neutralized particles can recycle, be retained by the plasma-facing components, or be exhausted through the pumping ports behind the 10 divertor units. In attached divertor operation only approx. 4 % of the incoming ion flux in the divertor is collected through the pump gap into the sub-divertor. When transitioning into detachment, neutral channels along the targets open up5, increasing the collection to 12 %, but also decreasing the particle plugging from 85 % to 50 %. Our gas balance calculations allow quantification of the particle sources and sinks associated with the plasma-facing components6. Hα measurements reveal that 85% of the 5.2E+22 recycling particles per second ionize in the divertor region, while 15% recycle far away from the recycling surfaces in the main chamber 4.
Active fueling was conducted with multiple systems (divertor and main chamber gas injection, pellet, and neutral beam injection) to reach high performance plasmas and detachment. The fuelling efficiency of these systems varied and with up to 90% was highest for the NBI, pellets of up to 80% and the gas injection between 12% and 44%.
The gas injection system was successfully operated in feedback density control and offered reliable access to stable detached plasma states. During detachment, heat loads drop dramatically and particle loads onto the divertor targets decrease by 65%. The installed divertor cryo-pumps will allow improved pumping in the next campaign, and it is expected to allow for density control even with the upgraded NBI and pellet injection fueling, as well as the expected wall source.References:
1 Jakubowski et al 2021 Nucl. Fusion https://doi.org/10.1088/1741-4326/ac1b68
2 Schmitz et al 2021 Nucl. Fusion https://doi.org/10.1088/1741-4326/abb51e
3 Kremeyer et al 2022 Nucl. Fusion https://doi.org/10.1088/1741-4326/ac4acb
4 Winters et al 2021 PPCF https://doi.org/10.1088/1361-6587/abe39c
5 Feng et al 2021 Nucl. Fusion https://doi.org/10.1088/1741-4326/ac0772
6 Schlisio et al 2021 Nucl. Fusion https://doi.org/10.1088/1741-4326/abd63fSpeaker: Kremeyer Thierry (Max Planck IPP) -
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Impurity leakage mechanisms in the W7-X island divertor under experimentally relevant operational space
Optimization of divertor concepts for the 3D stellarator boundary is still only in its beginning stages. Multiple requirements must be satisfied for a properly functioning divertor, including optimized particle exhaust for density control, significant power exhaust to ensure the survival of plasma-facing components, and sufficient impurity retention in the scrape-off layer (SOL) to avoid undue contamination to the confined plasma. Impurity retention in the SOL serves two other important aspects of divertor operation: 1. For a given number of impurity particles, there is a greater concentration of impurities near the pumping area, allowing for efficient exhaust (critical for Helium ash removal) and 2. it provides a larger density of seeding impurities for a given puff rate in the SOL, where they more efficiently radiate, improving power exhaust. Thus, understanding the driving factors for impurity leakage in current stellarator divertor designs is an important step towards developing better optimization criteria for future stellarator reactor divertors.
The EMC3-Eirene code was utilized to study the dominating transport mechanisms for impurity leakage in the Wendelstein 7-X island divertor under experimentally relevant density operational space, in the absence of drifts. The results are consistent with previous work indicating a transition to friction force dominance over the majority of the island SOL, at separatrix densities below those that were typically achievable in the previous operational campaigns[1][2]. The suppression of the ion thermal force is so strong that no impurity neutrals ionize beyond the parallel impurity flow stagnation point. Therefore, parallel transport does not play any direct role in impurity leakage from the island SOL in the simulations. Rather, it was found that the impurity density at the last closed flux surface (LCFS) was strongly sensitive to the anomalous cross-field diffusion coefficient assumed, indicating that perpendicular transport plays a strong role in limiting the impurity retention of the island divertor[3]. The main location of impurity leakage via the perpendicular transport channel appears to be the island O-Point region. The impurity dwell time of this location is rather long, owing to the low internal island field line pitch in combination with impurity flow stagnation in this location. This provides enough time for the impurities to diffuse perpendicularly to the LCFS.
The results indicate the importance of the proximity of the island O-Point to impurity sources (plasma-surface interaction, seeding sources) in limiting the impurity retention of the island SOL. Such a result, if validated experimentally, provides valuable information for design of future island divertors. Additionally, the results indicate that experimental measurements of impurity retention may provide a useful constraint on the value of the assumed anomalous cross-field diffusion coefficient used in simulation. Further work is needed to quantify drift effects[4], which may impact the strength of the effects discussed here[5].References
[1]Y. Feng et al, Nucl. Fusion 49 (2009) 095002
[2]Y. Feng et al, Plasma Phys. Control. Fusion 53 (2011) 024009
[3]V. R. Winters et al, Nucl. Fusion (submitted)
[4]K. C. Hammond et al, Plasma Phys. Control. Fusion 61 (2019) 125001
[5]D. M. Kriete et al, in preparationSpeaker: Victoria Winters (Max Planck Institute for Plasma Physics)
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Discussion Session: SOL Physics Board Room A
Board Room A
IAEA Headquarters
Convener: Anthony Leonard (USA) -
12:35
Lunch Board Room A
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IAEA Headquarters
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PFMC (part I) Board Room A
Board Room A
IAEA Headquarters
Convener: Richard Pitts (ITER Organization)-
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Overview of DEMO Divertor Architecture Design options
The divertor is one of the key components of the EU-DEMO reactor. The development of a reliable solution for the power and impurity particle exhaust is recognized as a major challenge toward the realization of DEMO. The pre-conceptual design activities for the EU-DEMO divertor are carried on considering two project areas: the ‘Target development’, focusing on the design of the vertical targets directly facing the plasma, and the ‘Cassette design and integration’, dealing with the design of the cassette structure and the integration of sub-components. The essential aim of the project is to develop in both project areas advanced design concepts for a divertor system capable of meeting the physical and system requirements defined for the EU-DEMO reactor.
In this work, a general overview of the EU-DEMO divertor cassette design is presented, considering systems requirements, structural assessments and interfacing systems. The design solutions adopted for the integration of the main divertor sub-components are described, in terms of layout and attachment to the cassette body (CB) of the Plasma Facing Components (PFCs), liner, reflector plates and cassette-to-vacuum vessel fixation system (nose at the inboard and wishbone at the outboard).
Different materials are integrated on the divertor cassette, requiring different cooling temperatures and leading to different behaviors to consider. In particular, Eurofer97 ferritic-martensitic steel has been selected for the cassette structure in order to meet the activation and radwaste requirements. As a consequence of this choice, in 2021 two different cooling circuits have been analyzed for the divertor cassette, one working at 180°C and 3.5 MPa as operating conditions for the Cassette Body, the other providing coolant water for the PFCs (CuCrZr pipes) at 130°C and 5 MPa.
In 2022 many critical remote handling issues have been studied such as the assembly and disassembly of the PFC, the radwaste material weight reduction, simplification of the piping layout and relative manufacturing, the optimization of the cooling condition between divertor and blanket using water under the same physical condition etc. These studies converged on an alternative solution using for the cassette and other Eurofer components (as reflector plates, shielding liner etc.) a cooling circuit with high pressure and temperature (p=15 MPa; T= 300 °C) similar to the breeding blanket cooling conditions.Speaker: Giuseppe Mazzone (aENEA Department of Fusion and Technology for Nuclear Safety and Security, via E. Fermi 45, 00044 Frascati, Italy) -
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Synergies in the technological developments of the W7-X and JT-60SA metallic divertor plasma facing components
For a successful operation of ITER and of future fusion reactors, several fusion devices are currently being exploited and upgraded. This paper focuses on the technical developments lead in JT-60SA (Japan) and Wendelstein 7-X (W7-X, Germany) fusion devices to provide support on the operation in metallic environment using operation conditions complementary to the current existing fusion devices.
The tokamak JT-60SA has been constructed in the framework of the broader approach with strong European support. After the operation of carbon actively cooled divertor, a transition to a metallic device is foreseen after 2033 for the “Integrated Research Phase II”. Tungsten actively cooled divertor target PFCs (W-ACD) will be installed to provide information on high-beta, inductive and non-inductive operations in metallic environment. The conceptual design of the JT-60SA W-ACD is planned to be achieved in 2026.
In parallel, Europe is also promoting studies on the stellarator concept, considered as the backup solution for the European reactor design. For this purpose, a transition of W7-X to a carbon free environment is planned. As for JT-60SA, the concept of the W7-X W-ACD with the related validated manufacturing process is planned to be qualified in 2026.
Analyzing the boundary conditions and loads related to the future W-ACDs of JT-60SA and W7-X, some similarities exist. First, these W-ACDs are planned to handle steady state thermal heat loads in the range of 10 MW/m². In both cases, the operation of actively cooled components based on carbon armour material (C-ACD) is planned before the transition to the W-ACD. Consequently, the boundary conditions (cooling conditions…) and interfaces of the already existing device dictate rather strongly the design options. The size of current C-ACDs for the two devices are also equivalent (~400 mm * 30-50 mm). To take benefit from these similarities, projects have been launched within the European work package called Divertor (WPDIV). The subprojects W7-X/JT-60SA will develop and demonstrate the feasibility of W-ACDs for W7-X and JT-60SA, within the boundary conditions and interfaces of the corresponding device, by full scale prototype manufacturing and high-heat-flux testing. In this paper, the concepts currently investigated for the metallic targets of JT-60SA and W7-X will be detailed and the rationale for the material and manufacturing processes choices will be presented.“This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.”
Speaker: Marianne Richou -
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Recent advances in EAST flowing liquid Li limiter experiments in support of divertor design for DEMO device
The investigation on effect of the flowing liquid Li limiter (FLiLi) on fuel and impurity, and heat fluxe during high confinement mode (H-mode) plasma was recently performed in EAST, aiming to provide an alternative resolution for the divertor design of a future DEMO device. Four generations of liquid Li limiters have been have successively designed and tested including a thin flowing film concept augmented by a thermoelectric magnetohydrodynamic (TEMHD) effect in generation 4[1-3]. The FLiLi was inserted into the edge in EAST plasmas with a limiter temperature of 300–400℃ and an auxiliary heating power of max. 8.3 MW. Analysis has shown that by using FLiLi, fuel particle recycling continuously decreased, and ≥80% of retained D particles were captured by the continual renewal of the Li redeposition film during FLiLi operation[4]. Thermal desorption performance of D2 in Li have been conducted to investigate possible Tritium removal in future device[5]. Impurity particle radiation, including high-Z and low-Z impurities, decreased significantly during the sequence of FLiLi operation. The decreased impurity radiation and recycling led to a modest increase (~10 kJ, 5%) in the plasma stored energy[6, 7]. Furthermore, H-mode could be easily access in an upper single-null configuration with ion grad-B drift away from the active divertor, reduce the L-H transition threshold power and gradually increase H mode duration (PRF~ 3.4 MW) while using the liquid Li limiter. Additionally, ELM mitigation was observed, accompanied by enhanced edge coherent MHD mode (ECM) and an edge harmonic mode, which was similar to that observed during boron powder and impurity gas injection in EAST [7]. With low magnetic field (Bt~1.9 T) and q95 ~ 4.2, by using liquid Li, large ELMs were effectively mitigated, i.e. increased ELM frequency and a decreased ELM size. Furthermore, vapor shielding effect due to an inserted FLiLi was investigated in the EAST device. The obvious decreased temperature and surface temperature oscillations of the liquid Li were observed with the presence of a Li vapor cloud during H mode discharges. A sudden Li burst probably due to Li droplet ejection from limiter surface produces a strong Li radiative band and decreases liquid Li surface temperature directly monitored by IR camera. Thermal analysis of the Li vapor indicated that the neutral Li vapor shielded >42% of the parallel heat flux, at the areas with relatively weak interaction. These results will provide a reference to design divertor with high heat flux in future reactors by allowing for a self-healing, self-replenishing surface with no susceptibility to neutron damage to partly ameliorate lifetime and power-exhaust issues of PFCs.
[1] G. Z. Zuo, et al., Nucl. Fusion 59, 016009 (2019).
[2] D. N. Ruzic, et al., Nucl. Fusion 51, 102002 (2011).
[3] J. S. Hu, et al., Nuclear Materials and Energy 18, 99 (2019).
[4] C. L. Li, et al., Plasma Phys. Control. Fusion 63 (2021).
[5] L. Li, et al., Nuclear Materials and Energy 28 (2021).
[6] G. Z. Zuo, et al., Phys. Plasmas 27 (2020).
[7] G. Z. Zuo, et al., Physica Scripta T171 (2020).Speaker: Dr Jiansheng hu (Institute of Plasma Physics, Chinese Academy of Sciences) -
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Proposal for an high Z liquid metal divertor
Power exhaust is a key mission in the roadmap to the future fusion reactor. Several alternatives have been proposed, among which the use of liquid metals (LM) as plasma facing materials. The presentation will thus focus on the latter, as it will give an overview on the proposal for an high Z liquid metal divertor. A review of the main experiments performed in the last few decades on the Frascati Tokamak Upgrade (FTU), as well on other fusion relevant devices in the word will be given.
The analysis of the results of the above mentioned experiments, have lead to the design of a tin based liquid metal target compatible with the actually foreseen DEMO plasma scenario.
Several aspects of the use of LM in a tokamak – such as evaporation, sputtering, retention, etc. – will be addressed since they have been a key parameters in the design of a LM in vessel components. In this reference, corrosion and compatibility issues will be addressed too.
To conclude, the design compatibility of the test Liquid Metal Divertor (LMD) module with the Divertor Tokamak Test facility (DTT) [1] will be shown. The possibility to test a liquid target in one of the four divertor test module in the DTT facility has been evaluated and will be presented in the framework of the DTT Research Plan.[1] G. M. Polli, “Divertor Tokamak Testing Facility (DTT): A Test Facility for the Fusion Power Plant”, (2021) Offshore Mediterranean Conference and Exhibition, 978-88946678-0-6
Speaker: Matteo Iafrati (ENEA)
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15:30
Break Board Room A
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IAEA Headquarters
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Poster Session II (all tracks) Board Room A
Board Room A
IAEA Headquarters
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Fabrication of CFC/Cu flat type Plasma Facing Components of HL-2M advanced divertor
HL-2M is a new tokamak machine in SWIP，featuring a samll angle slot (SAS) divertor and can perform advanced divertor configurations, such as snowflake and tripod. It is estimated that the wall loads on the target plates can be changed from 3 to 7 MW/m2 depended on normal and snowflake or tripod divertor configurations. In order to meet the operation requirements of HL-2M tokamak machine, an active-cooling design of divertor components is considered using CFC/CuCrZr plate mockups by brazing joining. The CFC is CX-2002U from Toyo Tanso Co., Ltd. Japan, which has very high thermal conductivity in the direction of fiber weaving. High heat flux tests (1000 cycles at a heat flux of 10 MW/m2) of small scale CFC/CuCrZr mockups in EMS-60 have proven that this fabrication process is feasible.
This report will introduce the divertor components manufacturing process, including the materials preparation, brazing technology, manufacturing, overall assembly and examination. Materials preparation mainly includes CFC completes surface metallization (modified by using Cr) and subsequent pure copper casting, bending and drilling of internal waterways of CuCrZr alloy. Low temperature brazing (<700℃) technology was performed to ensure the mechanical performance of CuCrZr alloy after brazing. Many new technologies have been developed and applied to the manufacturing process to ensure the accuracy and reliability of components. A variety of non-destructive testing technique (ultrasonic and 3D ultrasono-graphy) were applied in the inspection process to ensure the reliability of the welding interface. Positive and nigetive pressure helium leak detection ensure that cooling pipes do not have air tightness problems. All units have been manufactured by the optimized process and passed gas tightness and dimensional accuracy examinations, installation of the components on the HL-2M is in progress.Speaker: Fan Feng -
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Divertor leg length considerations for testing detached divertor operation compatibility with a high-performance core plasma
Divertor leg length requirements for testing divertor detachment compatibility with a hot X-point plasma and robust H-mode pedestal is examined with DIII-D data and modeling. Poloidal Te gradients consistent with this requirement are found to be determined by convective energy transport and particle balance constraints. These considerations provide estimates of divertor leg length requirements for existing and planned facilities that aim to test a Fusion Pilot Plant (FPP) divertor design and operation for issues such as integration of divertor heat flux control with core plasma operational scenarios.
Measurements from the divertor Thomson scattering (DTS) diagnostic on DIII-D indicate average poloidal electron temperature gradients of up to 200 eV/m through the dissipative region of a detached outer divertor leg with the radiating front mid-way up the leg. These gradients imply a minimum leg length of 15-20 cm in DIII-D for the dissipative region, Te ≤ 20 eV, to remain below the X-point. The spatial extent of the radiating region implied by the measured gradient is confirmed with visible imaging of the carbon impurity radiating state.
The observed divertor poloidal Te gradients are found to be consistent with convective energy transport through the divertor dissipative region, Te ≤ 40 eV, while transport dominated by conduction would imply much steeper gradients than observed experimentally. For transport dominated by convection, the characteristic Te poloidal scale length can be estimated as $L_{Te}\approx T_e /(dT_e /ds_{pol} ) \approx v_{pol} q_e T_e / L_z f_z n_e$ where $v_{pol}$ the effective plasma poloidal velocity from parallel convection and $ExB$ poloidal flow, $L_z$ is the radiative loss parameter and $f_z$ is the radiating impurity fraction. For carbon impurity radiation peaking near 10 eV, $n_e \sim 2x10^{20} m^{-3}$, a measured impurity fraction of $\sim 0.2%$ and $v_{pol}$ consistent with Mach 1 parallel flow, the calculated Te decay length is similar to that observed experimentally. UEDGE 2D-fluid simulation with drifts included of a detached divertor with 3 MW of input power and radiation from sputtered carbon reproduces most aspects of this conceptual model, including a poloidal Te gradient of ~250 eV/m and poloidal energy transport dominated by convection from both parallel flow and $E×B$ poloidal drift.
The implications for DIII-D divertor upgrades and future tokamak divertor design will be explored. For DIII-D high performance Advanced Tokamak (AT) scenarios with $≥$ 10 MW of injected power and a requirement for maintaining the X-point region at $T_e ≥$ 80 eV robust pedestal pressure, a divertor leg length of $≥$ 50 cm is indicated. This is a factor of 2-3 longer than the typical DIII-D configuration and would present a target for future divertor upgrade design. For future tokamaks the scaling appears to be favorable with larger size, higher field and higher density, but will also depend on the scaling of radial transport and plasma drifts.
Work supported in part by the US DOE under contracts DE-AC05-00OR22725, DE-AC52-07NA27344,and DE-FC02-04ER54698.Speaker: Anthony Leonard (USA) -
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Target technologies R&D for EU-DEMO divertor
Within the roadmap of the EU-DEMO reactor design, the divertor design is being developed in the framework of the EUROfusion Consortium Work package “Divertor”, sub-project DEMO (WPDIV-DEMO).
The ultimate objective of this sub-project is to deliver at least one holistic design concept and feasible technology options for the DEMO divertor and limiter eligible for the subsequent Engineering Design Activities (EDA) phase.
For the divertor targets, the qualified design and technology is that of ITER: monoblocks in W equipped with a copper interlayer of 1 mm thickness, obtained by casting or HIPing, welded by diffusion bonding or by brazing on a CuCrZr pipe. In parallel, for risk mitigation, dedicated R&D shall be conducted for alternative technologies with the aim of increasing the resistance of the components to damage due to neutron radiation.
Due to the high thermal gradients that are generated in the target (less than 10 mm elapse from 2000 ° C of the surface to the cooling water temperature), the heat sink material must have high mechanical strength and high conductivity. So, the CuCrZr is the best candidate. However, in addition to not being a low-activation material, which makes it unattractive for its use in fusion plants for the commercial production of energy, it undergoes a high degradation of its mechanical characteristics under the neutron radiation provided in DEMO. To limit the damage the CuCrZr should be kept at a temperature higher than 200 ° C, but not higher than 350. The narrow operating window is making the thermo-hydraulic design of the DEMO divertor complicated.
In particular, the R&D activities are aimed at the production of components with cooling composite pipe in W fiber and copper matrix. The W, present in a high percentage in these tubes, ensures mechanical strength, high conductivity, low neutron damage and less activation. Some promising results were obtained in the final phase of Eurofusion Horizon 2020, and up-scaling to larger components is the main object of the work in recent years.
In addition to the manufacturing of the pipes, it is necessary to develop and verify a reliable joining technology with W monoblocks. Joining techniques by brazing and by copper casting are being studied and tested. This work reports on attempts and results, not always encouraging, achieved.Speaker: Selanna Roccella (ENEA Department of Fusion and Nuclear Safety Technology) -
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Time-dependent SOLPS-ITER simulation for actuator design and system identification
Time-dependent SOLPS-ITER simulations have been extensively applied to various problems such as actuator design, feedback control through system identification, and physical interpretation of dynamic problem. Most SOLPS-ITER simulations focus on steady-state solutions, and the EIRENE module, a neutral solver, couples in a time-independent mode assuming quasi-relaxation within time step of 1e-3 sec, which is several orders of magnitude larger than the fluid plasma time step. We made the neutral time step the same as the plasma time step and performed a full time-dependent simulation considering the neutral census data. This feature is applied to the design of SPARC’s Louvre structure [1], which controls divertor plasma parameters by regulating neutral conductance from the divertor to the pump. The response of the plasma and neutral parameters can be captured with the timescale that enables us to design the actuator considering time-dependent control capability. Feedforward SOLPS-ITER simulations were carried out in KSTAR using the recently developed time-dependent gas puff feature, and system identification based on these data was carried out with DMD and SINDy. Through this reduced model, feedback control of upstream density was performed under IPS environment. The target flux bifurcation observed in the KSTAR experiment is a dynamic problem, and a time-dependent SOLPS-ITER simulation was performed for the physical interpretation. It was found that X-point radiation was the main cause, and the bifurcation time scale observed in the experiment was also reproduced in the simulation within a factor of 2. The validity of data-driven system identification and feedback control, including bifurcation, is also tested.
Work supported by US-DOE under contract DE-AC05-00OR22725, by the ORNL Laboratory Directed Research and Development Program, and by the Innovation Network for Fusion Energy (INFUSE).
[1] A. Q. Kuang, et al. J. Plasma Phys 86.5 (2020)
Speaker: Jaesun Park (Oak Ridge National Laboratory) -
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Early Observations of Divertor Detachment on MAST Upgrade
Future large, high power fusion devices such as ITER and its successors will require strong dissipation of the plasma energy, momentum and particles reaching open field lines before reaching the surfaces of the divertor; detachment will play a dominant role in that dissipation. This contribution presents initial observations of divertor detachment from the MAST Upgrade spherical tokamak in conventional and Super-X divertor configurations, emphasising the physical processes governing the onset and evolution of detachment.
In the first physics campaign intensive studies of Ohmic discharges with plasma current of 600kA and 750kA were performed in a double null topology. In experiments where the core density increases continuously, the roll-over in the outer divertor ion saturation current is coincident with the surface heat flux dropping to levels close to the detection limit of the infrared cameras, in both conventional and Super-X divertor configurations. In the conventional divertor configuration, the roll-over in the outer divertor ion flux is driven by a reduction in both the power entering the divertors and upstream plasma pressure. Analysis of passive spectroscopy measurements of atomic deuterium Balmer and molecular Fulcher band emission suggest that the ion flux rollover is caused by a reduction in the divertor ionisation source. The onset of detachment in the Super-X configuration occurs when the line-average density is ~10% of the Greenwald density limit, 50% lower than in a conventional divertor configuration. These observations are supported by visible imaging measurements of Balmer and Fulcher emission withdrawing from the divertor target with increasing upstream density, and will support more quantitative analysis of the divertor conditions. Divertor spectroscopy and Thomson scattering suggest that the divertor Te during most of the detached phase is very low, 0.1 < Te < 0.5eV.
This contribution will present a synthesis of these initial observations of divertor detachment on MAST Upgrade, including a description of novel data analysis techniques including Bayesian approach to estimate the upstream density from mid-plane Thomson scattering profiles and analysis of divertor spectroscopy data. A comparison with SOLPS-ITER modelling will be presented, showing very good agreement with the results from the Super-X configuration, including detachment threshold and divertor fluxes.Speaker: James Harrison (United Kingdom Atomic Energy Authority) -
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A Continuous V Shape Water Cooled Divertor Structure: Flexible with Plasma Configurations and Improved Heat Flux Capability
Aiming at achieving steady state operation of Tokamak, the divertor is supposed to exhaust particle and heat flux. Plasma configuration is one of the most important drivers for divertor design determining the position of stricken point and heat flux intensity. In order to provide flexibility with plasma configurations, a continuous V shape divertor structure with consistent heat removal capability is proposed with tolerance on variation of stricken point location. Taking into account the gradually decreasing heat flux intensity from stricken point, the divertor target is divided into three zones which are high, medium and low heat flux zones. The high heat flux zone covers stricken zone variation range in which the maximum heat removal capability of plasma facing unit is 15 MW/m2. The W/Cu hypervapotron, as plasma facing unit, is designed and optimized with sufficient heat transfer performance. The explosive bonding technology is tested and applied to produce the continuous V shape structure made of heterogeneous materials.
Speaker: Xuebing PENG -
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An overview of the conceptual design of the plasma-facing components of the DTT divertor
The Divertor Tokamak Test facility (DTT) [1] is a fusion device currently under construction in ENEA Frascati. Its main scientific goal is to investigate advanced solutions for the heat exhaust in future fusion power plants, such as DEMO. A key step in the success of DTT is the development of integrated plasma scenarios where good core performance is achieved together with acceptable conditions at the wall. For this reason, a substantial number of different magnetic configurations are meant to be tested in DTT. Such a flexibility requirement is the central drive of the design of the first DTT divertor, which must accommodate a multitude of strike points, generally located at various positions accordingly to all the equilibria. Moreover, a proper length of the divertor legs must be ensured to provide both a significant divertor volume to increase the radiated power at the edge, and additional spreading of the energy due to the turbulence transport below the X-point. Such high level of flexibility makes the design of the DTT divertor a first of its kind.
In this contribution, we present an overview of the conceptual design of the plasma-facing components of the first DTT divertor. The poloidal profile of the divertor has been chosen to accommodate the following magnetic configurations (in order of priority): Single Null, X-Divertor (XD), Negative Triangluarity and Snowflake divertor. The plethora of possibilities lead to a wide divertor design, which provides a substantial volume and can host different shapes of divertor legs. The Plasma Facing Units (PFUs) of the divertor are virtually entirely covered by tungsten monoblocks, to allow for different locations of the strike points, while whitstanding steady state heat loads up to 20 MWm-2. Each PFU is made of three segments, Inner Vertical Target, Dome and Outer Vertical Target, connected hydraulically in series, resulting in a single unit connecting the high field side to the low field side. The use of shielding plates are avoided by ensuring, through proper design of the central PFU segment, that the inlet/outlet naked pipes are in the magnetic shadow of the Dome. The Dome, which acts in all respects as a target, is flat in order to minimize its height, thus allowing a longer leg for the equilibria having the outer strike point impinging on it (e.g., the XD). The technology chosen for the PFU manufacturing is the one developed in the ENEA laboratories and qualified for the fabrication of the ITER Inner Vertical Targets. The main results from numerical computation will be reported, mainly concerning the structural and hydraulic verification of the assembly.
[1] G. M. Polli, “Divertor Tokamak Testing Facility (DTT): A Test Facility for the Fusion Power Plant”, (2021) Offshore Mediterranean Conference and Exhibition, 978-88946678-0-6Speaker: Giacomo Dose (University of Rome “Tor Vergata”) -
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Divertor Detachment In Negative-Triangularity Configurations In The TCV Tokamak
Experimental observations on TCV [1] and DIII-D [2] have shown that negative triangularity L-Mode discharges can exhibit H-mode grade confinement, opening the possibility for high confinement reactors that side-step the challenges associated with H-mode such as ELMs, narrow scrape-off layer widths, and density control. To ensure safe power exhaust that protects the plasma facing components, partially or fully detached divertor operation will still, however, be required. This work, therefore, investigates detachment of TCV ohmic L-Mode negative triangularity (NT) configurations and compares them to similar positive triangularity (PT) cases. Detachment is generally found harder to attain in NT, where sufficient cooling (< 5eV) of the outer target is not achieved in core density ramps and, with N2 seeding, only at the cost of confinement degradation. While changes in connection length and divertor shape were initially thought responsible, experiments with matched poloidal outer leg length and effective connection length, or matched divertor geometry, still show an increased difficulty in reaching detachment for NT, seen by reduced outer target cooling and no clear movement of the CIII front towards the X-point. Discharges with matched divertor geometry but changes of the top triangularity indicate a generally lower divertor neutral pressure in NT plasmas, associated with a lower D2 flux required to achieve a similar core density ramp, hinting at a difference in particle confinement. This contribution will also explore the role of lambdaq, previously measured to be smaller in L-Mode NT than in L-Mode PT [3]. Overall, this study indicates that, while NT represents a promising solution towards ELM-free, high confinement scenarios, the core-edge integration remains challenging.
[1] Y. Camenen et al, Nucl. Fusion 47 510 (2007)
[2] M. E. Austin et al, Phys. Rev. Lett. 122, 115001 (2019).
[3] M. Faitsch et al, Plasma Phys. Control. Fusion 60 045010 (2018).Speaker: Olivier Février (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland) -
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Divertor heat flux mitigation with boron and boron nitride powders in DIII-D
Injection of boron (B) and boron nitride (BN) in powder form into the upper closed divertor at DIII-D showed a substantial drop in divertor electron temperature from 30 eV to below 5 eV, increase in divertor neutral compression by up to an order of magnitude, and transition into stable detachment [1]. A decrease in wall fueling, main chamber neutral pressure, and the reduction of oxygen, carbon, and neutral deuterium line emission in the plasma edge and divertor demonstrated additional conditioning of the main plasma-facing components caused by the cumulative injection of the low-Z non-recycling materials.
Real-time wall conditioning with B and BN powders has been demonstrated before at AUG and DIII-D [2-4]. Numerical modeling suggests that boron is, in addition to nitrogen, very suitable for reducing divertor peak heat loads at relatively low concentration and low core contamination [5]. To investigate this experimentally, B and BN powders of 65-150 um particle size were injected into the DIII-D closed small-angle slot divertor [6] in upper-single-null ELMy H‑mode plasmas (Ip~1 MA, BT=2 T, PNBI=6 MW, fELM=80 Hz, <ne>=3.6-5.0x1019/m3) in 2-s intervals at rates of 3-200 mg/s. BN powder at 50 mg/s showed transition into detachment while rates of 200 mg/s triggered n=2 tearing modes and a drop of up to 24% in energy confinement and a reduction in neutron rates. Plasma fluid simulations coupled with dust transport modeling were used to study the effect of the powder particle size on ablation and impurity migration in the divertor and scrape-off layer. The results show that larger powder particles can escape the divertor before they fully ablate, resulting in less localized boron fluxes advantageous for conditioning and more uniform power losses in the plasma boundary.
The experimental and modeling data suggest that boron and boron nitride injection provide an alternative approach for safe divertor power exhaust. Previous experiments have shown that using the BN powder impurity mix reduces the ammonia content by 90% compared to N2 gas injection [3]. Furthermore, real-time injection of B and BN powders brings the additional benefit of reduced recycling, better density control, and suppression of intrinsic impurity sources due to active conditioning of the divertor plasma-facing components during plasma operation. Thereby, powder injection could at least supplement conventional methods of impurity seeded power exhaust and wall conditioning, affecting the divertor integrity, performance, and operability of a fusion pilot plant.[1] F. Effenberg et al, submitted to Nucl. Fusion 2022, arXiv:2203.15204
[2] A. Bortolon et al 2019 Nucl. Mater. and Energy 19 384-389
[3] R. Lunsford et al 2019 Nucl. Fusion 59 126034
[4] A. Bortolon et al 2020 Nucl. Fusion 60 126010
[5] A. Yu. Pigarov et 2017 Phys. Plasmas 24, 102521
[6] H.Y. Guo et al 2019 Nucl. Fusion 59 086054This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-AC02-09CH11466, DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725.
Speaker: Dr Florian Effenberg (Princeton Plasma Physics Laboratory, Princeton, NJ, 08543, USA) -
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ELM Mitigation Enabled by Control of Neutral Recycling with New EAST Divertor
Divertor recycling control is enabled by a new tungsten lower divertor on EAST, featured by a right-angled closed corner joining the vertical and horizontal target plates. ELM mitigation is observed as the outer strike point moves from the vertical target to the horizontal target with a significant reduction of the pedestal density gradient. Furthermore, the new closed corner divertor exhibits significantly better recycling control, power handling, lower detachment threshold and core confinement compatibility than the conventional vertical-target divertor.
FIG. 1. SOLPS-ITER simulation results for the 3 different divertor strike point locations with full drift effects. The two-dimensional distributions of particle ionization source SD+ for the strike point located on (a) the left side (further away from the corner) and (b) the middle of the horizontal target, and (c) the vertical target.
As shown in Fig. 1, SOLPS-ITER simulation indicate that when the strike point is located on the horizontal target most particles from the upstream SOL flow into the outer divertor slot and hit the horizontal target plate and the ionization source appears to concentrate in the vicinity of the horizontal target, due to the trapping of the recycled particles in the closed corner area, leading to a reduced pedestal fueling from the lower divertor, thus reducing the pedestal density gradient and elevating the SOL density. In contrast, when the strike point is located on the vertical target, a much stronger ionization source appears in the vicinity of the X-point, especially inside the separatrix [Fig. 1(c)]. The vertical target plate reflects recycled neutral particles towards the private region where the electron density and temperature are much lower than the SOL. Therefore, the neutrals cannot be fully re-ionized in the vicinity of the X-point and tend to diffuse into the pedestal. Furthermore, as the baffle area is closer to the divertor strike point for the vertical target case, much more particles from the upstream SOL hit the baffle, which also enhance the pedestal fueling. This results in a much steeper density gradient and a lower density in the SOL for the strike point on the vertical target.
This paves a new path in ELM mitigation through tailoring pedestal structure with divertor condition control. These new results may have strong implications for future fusion reactors where a low pedestal density gradient is anticipated due to much higher neutral opacity and lower pedestal fueling.Speakers: Mr Guosheng Xu (Institute of Plasma Physics, Chinese Academy of Sciences), Mr Xin Lin -
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Modeling a Lithium Vapor Box Divertor and Resulting Ion Flows on NSTX-U using SOLPS
Divertor detachment with medium-Z impurities seeded through gas puffing can entail radiating regions within the last closed flux surface. The lithium vapor box divertor seeks to detach via near-target low-Z lithium evaporation with the result that such a radiating region does not form. We show SOLPS-ITER predictions for the effect of a lithium vapor box divertor on NSTX-U. Past work has shown the lithium vapor box is capable of reducing 65 MW/m$^2$ of perpendicular heat flux to below 5 MW/m$^2$ for upstream lithium densities equal to a few percent of the electron density. Those results are tested for their sensitivity to choices of transport coefficients, recycling coefficients and puffing location. Even when transport coefficients are reduced to provide less particle flow from the core and higher heat flux at the target, sub-10 MW/m2 solutions are available to the lithium vapor box, as compared with an unmitigated perpendicular heat flux of 92 MW/m$^2$. Private Flux Region (PFR) fuel puffing is seen to be more effective at reducing upstream lithium content while Common Flux Region (CFR) fuel puffing is seen to be more effective at heat flux reduction. The efficacy of both puffing locations is increased by increases to the divertor recycling coefficient. Increased recycling at walls upstream of the baffles improves the effect of the puffs, leading to cases with lower upstream lithium content for less heat flux.
Speaker: Eric Emdee (Princeton Plasma Physics Laboratory) -
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Numerical simulation of neutron flux distribution for DEMO Divertor heterogeneous model concept
Results from calculations of the neutron flux in the divertor cassette body for the Helium Cooled Pebble Bed and Water-Cooled Lithium Lead concepts are presented in this paper. For both cases under investigation, the same divertor model setup and designated DEMO neutron source were utilized. Neutron transport calculations were performed using the MCNP6 and the FENDL-3.1 nuclear data library. The ADVANTG (AutomateD VAriaNce reducTion Generator) tool and MCNP code, which were utilized for the variance reduction and particle movement appropriately, are also included in this study's estimations of the neutron-induced dose rate. The development of the dose rate and neutron flux maps for the DEMO reactor's divertor was made possible by the use of such a coupled computing technique.
Results from the study of neutron flux calculations in the divertor cassette body were considered using in HCPB and WCLL concepts as a breeding blanket (BB). To obtain the presented results in this paper, the MCNP6 code was combined with the ADVANTG code and FW-CADIS variance reduction settings, which allowed for less computational time and statistical error. When employing the WCLL BB model, the statistical error fell by an average of 1.13 and 2.1 times for neutron and dose rate estimates, compared to 3.1 and 5.4 when using the HCPB BB concept design. Calculations of the neutron flux in various cells of the MCNP model revealed that the two examined BB models had similar neutron distributions in the energy range from 0.794 MeV to 15.6 MeV. Since no neutrons are created in the EU DEMO utilizing DD or DT fuel cycles above 15.6 MeV energy, there are no neutrons above this energy threshold. The average number of neutrons in the WCLL seemed to be 1.29 times higher than in the HCPB in the lower energy zone (i.e., 1.05e-11 MeV - 0.794 MeV).
As stated in the abstract of the paper, the effect of the variance reduction tool—such as the WW produced using ADVANTG—was evaluated. The ratio of the calculation results can reach up to 1.28 for WCLL and 1.80 for HCPB BB instances when utilizing WW and without WW, according to the neutron flux maps. Additionally, the effect of WW on dose rate estimates is comparable; the ratio rises by 3.46 and 3.90 times, respectively for HCPB and WCLL BB at investigated locations that are furthest from the source below the divertor.Speaker: Gediminas Stankunas (Lithuanian Energy Institute) -
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Tungsten sources and core contamination in WEST diverted configurations
Over the first phase of WEST exploitation, the full tungsten environment consisted in a mixture of coated tiles and actively cooled ITER-like monoblocks. Using a combination of LHCD and ICRH heating, diverted L-mode scenarios were extended up to 50 s and stationary heat loads reaching 6 MWm-2 were deposited on ITER-like monoblocks. Despite input power levels generally above expected thresholds for onset of improved confinement modes (3-4 MW), unmitigated core radiation levels, in the range of 50% of the input power, prevented routine access to those improved modes. Core tungsten concentrations are generally estimated in the range of 10-4 from absolutely calibrated VUV spectroscopy, bolometry or soft X-ray.
Tungsten sources from divertor targets do show significant variability across plasma scenarios despite relatively constant core radiation levels, suggesting efficient screening even in compact divertor geometry. On the contrary, RF power coupling requires to adapt the plasma-antenna gap, in turn potentially enhancing the source strength and penetration from their tungsten limiters. These limiter sources are believed to rule core contamination, despite that there is no strong experimental evidences supporting this statement. Introduction of boron nitrite bumpers in the tokamak chamber helped stabilizing the plasma start-up (initially done on tungsten bumpers) but did not show hysteresis effects on diverted scenarios. The choice of replacing RF antennae limiters by boron nitrite coated tiles shall be taken next year, although this moves WEST away from a full metal environment, and despite no strong experimental arguments.
The true problem is rather to understand if the constant 50% radiation level in WEST can effectively be tamed by replacing some or all antennae limiters, or if this level of radiation is simply a stiff balance between power exhaust and core radiation. Based on SOLEDGE & ERO2 simulations of the tungsten transport at the edge of WEST [1], a phenomenological model can be built in order to relate core radiation and gross sources from different locations. Parametric sensitivity study help to disentangle the dominant mechanisms at task in the transport of the eroded tungsten to the core plasma. In particular, it is possible to recover a 50% radiation level across any density and input power variations, due to divertor sources. The model will be explored to test if divertor or antenna limiters have similar impact on global scenarios.[1] S. Di Genova et al, Nucl. Fusion 61 (2021)
Speaker: Dr Nicolas Fedorczak (IRFM/CEA)
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Convener: Rudolf Neu (Max-Planck-Insitut für Plasmaphysik)-
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Tungsten lifetime assessment for divertor component design
For the European fusion energy program, one of the main objectives is to establish a physical and technological basis for reliable power exhaust during the DEMOstrational power plant operation. As a consequence of the particles bombardment (ions, electrons, neutrals) coming from the plasma, the radiation, the energy conducted along the magnetic lines and from neutron irradiation, the divertor plasma facing components (PFCs) are the most thermally exposed. Considering these functional constraints, efforts are on-going to improve the material/component design and develop numerical codes which can achieve DEMO relevant Finite Elements (FE) simulations to assess tungsten damage and predict lifetime. Existing tools gave the opportunity to improve our understanding regarding the damage mechanism of such components under pure thermal heat flux. These simulations are representative to a number of existing fusion devices operational conditions but are not relevant for the DEMO operation. In that context, one goal of the T-REX project (funded by EUROfusion) is to provide for tungsten (used as armor material in the DEMO divertor PFC baseline) a FE modeling tool (based on ANSYS software) able to assess, at the macroscopic scale, the relevant mechanical stress and strain fields under DEMO operation conditions (plasma bombardment, conducted power, neutron irradiation).
The last T-REX developments are presented. For the first time, tungsten damage is assessed taking into account the influence of both isolated and combined heat flux/neutron loading on the tungsten thermal and mechanical properties (incl. the softening kinetics) and assess the Deuterium/Tritium (D/T) concentration profile considering the representative mechanical stress and strain fields estimated. Thermal cycling (10s ON/ 10s OFF) simulations are performed assuming a homogenous heat flux at either 10 or 20 MW/m² on the current DEMO divertor component design upper surface. Collected results show that thermal conductivity degradation (due to neutron irradiation) leads to temperature shifts (up to 400°C at 20 MW/m²). As expected, this trend promotes the tungsten softening (recrystallization/restoration). Results highlight that softening occurs after only twenty thermal cycles at 10 MW/m², which is not expected under pure thermal load (w/o neutron irradiation). Besides, collected results highlight that the tungsten damage mechanism (ductile/brittle) depends strongly on the neutron fluence (dpa) and on the tungsten microstructural state (softened or not) under irradiation. Lastly, D/T concentration profiles obtained for tungsten under pure heat flux at 20 MW/m² (w/o irradiation) are given. This first application highlights the influence of the plastic strain increment generated after each thermal cycle on the D/T trapped in tungsten (which gives trends useful for safety).Speaker: Alan DURIF (CEA) -
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Perspectives of chemical vapour deposition for the fabrication of tungsten fibre-reinforced composite components
Due to its unique property combination tungsten materials are the preferred choice for high-heat-flux-loaded areas in future fusion power plants. However, tungsten has a high brittle to ductile transition temperature and is prone to operational embrittlement due to high temperature and/or fast neutron irradiation. Tungsten fibre-reinforced tungsten composites utilize extrinsic mechanisms to improve the fracture toughness and thus mitigate this drawback. In these composites high strength drawn tungsten wire is used in combination with a tungsten matrix produced by chemical vapour deposition (CVD) or powder metallurgy (PM), respectively. Although little experience exists regarding the use of CVD for the production of bulk W materials this technique offers several interesting properties. In this contribution we will give an overview of the development of this production technique with a focus on recent results.
Using gaseous tungsten hexafluoride (WF6) together with hydrogen (H2), CVD allows the growing of solid tungsten on surfaces at temperatures above 350° C. Despite some success with a multilayer fibre preforms the focus in recent years has been on upscaling with a single-layer technique. Here, the upgrade both in textile techniques and CVD equipment allowed the fabrication of larger samples and the assessment of the size effect on mechanical properties. Recently, the use of yarns industrially produced from single W filaments has been established. These improved the reproducibility significantly.
However, the question of component design brought back the attention to more complex 3d structures recently. Although challenging the application of an infiltration process on textile preforms would allow a huge design freedom. The establishment of process understanding and modelling allowed to perform feasibility studies to assess the effect of process parameters like gas flow, local pressure and temperature. The combination with the tungsten fibre-reinforced copper concept offers a unique possibility for an optimized component design. Especially the improvement of joints, the load orientated fibre architecture and the possibility of larger structures are interesting. Further, the good resistance of the W fibres against irradiation embrittlement is an important benefit in view of future fusion reactor applications.Speaker: Johann Riesch (Max-Planck-Insitut für Plasmaphysik) -
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Plasma-facing components based on tungsten fiber-reinforced tungsten composites
Plasma-facing components based on tungsten fiber-reinforced tungsten composites
Y. Maoa,c, J.W. Coenena,e, J. Rieschb, X.Tanc, C.Chenc, A. Terraa, C.Liud, T. Höschenb, Y. Wuc, Ch. Broeckmannd, R.Neub,f, Ch. Linsmeiera
a Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Partner in the Trilateral Euregio Cluster, 52425 Jülich, Germany
b Max-Planck-Institut für Plasmaphysik, 85748 Garching b. München, Germany
cSchool of Material Science and Engineering, Hefei University of Technology, Hefei 230009, China
d Institut für Werkstoffanwendungen im Maschinenbau (IWM), RWTH Aachen University, 52062 Aachen, Germany
eDepartment of Engineering Physics, University of Wisconsin -Madison, Madison, US
fTechnische Universität München, 85748 Garching, GermanyFor the first wall of the future fusion reactor, unique challenges on materials in extreme environments require advanced mechanical and thermal properties. Tungsten (W) is the main candidate material for the first wall of a fusion reactor, as it is resilient against erosion, has high melting point and thermal conductivity, and shows rather benign behavior under neutron irradiation. However, the intrinsic brittleness of tungsten is a concern with respect to the fusion environment with high transient heat loads. Additionally, neutron induced effects e.g. transmutation add to embrittlement and are crucial to material performance. To overcome the brittle issue of tungsten, tungsten fiber reinforced tungsten (Wf/W) composites are being developed relying on an extrinsic toughing principle.
Recently, progress has been made for upscaling the production by powder metallurgy (PM) routes. Using large scale spark plasma sintering facility, samples with a diameter of 105mm and a thickness of 30 mm can be produced based on the porous matrix Wf/W concept, allowing the preparation of mock-ups (both flat tile design and monoblocks) for high-heat flux testing. Additionally, a new breakthrough has been made in terms of unidirectional long fiber Wf/W using powder metallurgy process. By alternately placing W weaves and W powders in the graphite mold, unidirectional long fiber Wf/W composite material has been prepared through PM process. The newly developed material shows a significant improvement regarding the mechanical properties, compared to the short fiber Wf/W. Apart from manufacturing progress, for the development of plasma-facing components, the issues regarding material joining, plasma erosion, tritium permeation, and fuel retention have also been tackled.Speaker: Dr Yiran Mao (Research Center Jülich) -
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Numerical design optimization of plasma-facing components using functionally graded materials
Divertor plasma-facing components (PFCs) of future fusion devices will have to deal with even more extreme heat loads than encountered at present. For ITER, it is expected that the current monoblock concept, where tungsten blocks are mounted on a copper cooling duct, will be sufficient. However, it is unsure whether these monoblocks can withstand the even more extreme conditions expected in future machines such as DEMO [1]. Therefore, solutions are required that either reduce the heat load or that improve the PFC design. Concerning the latter, several novel material concepts have arisen over the past decades, such as functionally graded materials (FGMs). These FGMs allow tailoring the local material properties in order to mitigate stress concentrations and therefore increasing the lifetime of the PFCs [2]. However, designing these components optimally in a manual way becomes cumbersome considering the large amount of degrees of freedom in the design, especially when additional design constraints have to be satisfied. For example, tungsten recrystallization and embrittlement has to be prevented by keeping the local temperature between the ductile-to-brittle transition temperature (DBTT) and the recrystallization temperature during operation wherever tungsten is present.
In this contribution, which is based on [3], we show how gradient-based numerical optimization methods can be used to quickly and automatically find the best possible design. A key ingredient here is the adjoint approach that allows all sensitivities to be computed with only a single additional simulation step, regardless of the amount of design variables. We apply this methodology to find the optimal material distribution in a W-Cu FGM-based ITER-like monoblock that minimizes the stress concentrations under steady loading conditions. At the same time, constraints on the local temperature are included to automatically prevent tungsten recrystallization and embrittlement for the operational scenario envisaged. To this end, a novel augmented Lagrangian strategy was employed to deal with arbitrary engineering constraints. We show that the optimized designs lead to significant stress reductions in the bulk of the monoblock, though the obtained design depends on model uncertainties, such as the stress-free temperature, and on the cost function formulation and constraints considered. For the latter, especially the DBTT of tungsten was shown to have a profound influence on the final design. In the future, characterization of and design for unsteady heat loads should be considered as well, given the importance of thermal cycling and fatigue due to for example fast transients and the pulsed operational regime.
Finally, we present an outlook on how the methodology can be applied for other design concepts, for example the design of first wall components, additively manufactured composites, different coolant channel geometries, shaping of the plasma-facing surface of the monoblocks, … Given that these techniques allow to quickly explore the potential of these novel designs in a cheap way, the design process can be significantly accelerated.
[1] J. H. You et al., Nucl. Mater. Energy 9, 171-176 (2016)
[2] J. W. Coenen et al., Phys. Scr. T167, 014002 (2016)
[3] S. Van den Kerkhof et al., Nucl. Fus. 61, 046050 (2021)Speaker: Sander Van den Kerkhof (KU Leuven)
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Conveners: Richard Pitts (ITER Organization), Rudolf Neu (Technische Universität München & MPI für Plasmaphysik) -
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Convener: Bruce Lipschultz (University of York)-
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Model advancements in mean-field plasma edge codes to enable computationally achievable simulations of the ITER and DEMO reactors
Plasma edge codes are the workhorse for divertor design for future machines. Models that correctly account for kinetic neutrals, anomalous transport, and detailed wall geometry are essential to extrapolate current knowledge towards reactors. Yet, the required highly collisional regimes and large size lead to inacceptable runtimes. We report on the status and remaining challenges of some recently developed models to address these issues.
To accelerate the simulations in high-recycling and detached conditions, advanced fluid neutral (AFN) and hybrid fluid-kinetic approaches for the neutral particles were recently developed. The AFN models significantly improve the results of classic fluid neutral models due to improved boundary conditions and transport coefficients, consistently derived from the underlying kinetic description [1,2].
The AFN models provide an alternative for a kinetic model in high-collisional regimes, but these conditions are usually only present in a small part of the simulation domain. We give an overview of several hybrid fluid-kinetic approaches with their advantages and drawbacks that were developed to address this. The most mature approach uses a spatial decomposition of the computational domain in a fluid and kinetic part, superposed with a condensation process that transfers atoms from kinetic to fluid populations in high-collisional regions [4]. The hybrid model for the atoms is coupled to a kinetic model for the molecules, resulting in hybrid-kinetic discrepancies that remains within 20% [4,5] while providing a factor 10 speed-up. Extended hybrid methods to further eliminate model errors are being explored [6,7].
Accounting for anomalous transport in principle requires fluid turbulence codes, but their cost will remain prohibitive for routine reactor simulations in the near future. The current ad-hoc treatment in edge codes through manually tuned transport coefficients leads to large model uncertainties. Moreover, experimental data to tune the coefficients is not available for future reactors. A more self-consistent description of this transport has recently been developed, relating the transport coefficients to the turbulent kinetic energy κ, which is solved from a corresponding transport equation [8]. Analysis of turbulence simulations suggests that the diffusion coefficients scale with the square root of κ. The κ-model has proven successful in capturing some key features as the ballooned nature of anomalous transport [9,10].
These models are part of the newly developed extended-grids version of SOLPS-ITER [11], paving the way towards efficient and accurate edge simulations at the reactor scale with this package.
[1] N. Horsten, et al., Nucl. Fusion 57 (2017) 116043
[2] W. Van Uytven, et al., Nucl. Fusion (2022)
[3] D. Borodin, et al., Nucl. Fusion, accepted
[4] W. Van Uytven, et al., Contrib. Plasma Phys. (2022) e202100191
[5] N. Horsten, et al., Nucl. Mater. Energy, submitted
[6] N. Horsten, et al., J. Comput. Phys 409 (2020) 109308
[7] B. Mortier, et al., SIAM J. Sci. Comput. 44 (2022) A720
[8] R. Coosemans, et al., Contrib. Plasma Phys. 60 (2020) e201900156
[9] S. Carli, et al., Contrib. Plasma Phys. 60 (2020) e201900155
[10] W. Dekeyser, et al., Contrib. Plasma Phys. (2022) e202100190
[11] W. Dekeyser, et al., Nucl. Mater. Energy 27 (2021) 100999Speaker: Prof. Martine Baelmans (KU Leuven, Department of Mechanical Engineering) -
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Self-consistent integration of plasma transport and divertor physics in the SOLEDGE3X-EIRENE code: status, results and prospects
Heat and particle exhaust in tokamaks is determined by a complex interplay between plasma transport processes, including turbulence, and a set of physical phenomena due to the interaction between the plasma, the solid wall, recycling or injected neutrals and intrinsic or seeded impurities. The comprehensive modelling of the physics at play requires a consistent integration of each of these mechanisms in a single numerical tool together with a realistic description of both the magnetic and the wall geometries. Due to the complexity of such task, the state of the art has for long been divided between mean-field codes, lacking a self-consistent description of transverse transport, and turbulence codes, ignoring most of the other aspects of the physics and most often in simplified geometries. Nevertheless, significant progress has been made in the last few years with 3D turbulence codes reaching the capability to run in realistic magnetic geometries [1]. The integration in these tools of the rest of the relevant physics is currently the focus of numerous teams in the community.
In this contribution, we report on such effort with a review of recent progress in the development and applications of the SOLEDGE3X-EIRENE code package [2]. The main features of the code are presented together with the numerical solutions that were adopted to address specific issues.
The SOLEDGE3X plasma solver solves fluid equations based on collisional closures. Both the standard Braginskii and the Zhdanov closure have been implemented and are showed to coincide for simple hydrogenic plasmas. Using the Zhdanov closure, multi-component plasmas with an arbitrary number of ion species can be modelled, each with individual temperatures and possibly at non-trace concentrations. Neutrals can be included either via fluid neutral models embedded in the plasma solver or via a coupling to the EIRENE Monte-Carlo kinetic code. Specific schemes have been developed to improve the robustness and efficiency of simulations with kinetic neutrals up to reactor relevant detached plasmas. Electrostatic fluid drifts can be activated on request.
A key asset of the code is its geometrical flexibility. The option is offered to run simulations in 2D or in 3D in arbitrary toroidally magnetic configurations, either limited or with an arbitrary number of X-points. This is in practise made possible by the use of a structured flux-surface-aligned mesh combined with a domain decomposition. The wall shape is arbitrary, 2D or 3D, and imposed through the use of mask functions inspired from penalization methods. We discuss the pros and cons of these discretization choices that were made as a compromise between contradictive demands between turbulence and divertor modelling.
Finally, the capabilities offered by the code are illustrated by recent application results on medium sized tokamaks (WEST, TCV, AUG) as well as on ITER relevant cases.Speaker: Patrick Tamain (CEA Cadarache)
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Convener: Bruce Lipschultz (University of York)-
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Parametric scaling of power exhaust in EU-DEMO SOLPS-ITER simulations
Investigations of scaling of power exhaust processes in the SOLPS-ITER database for single-null, Super-X, and X-divertor configurations of the EU-DEMO are conducted and compared to predictions based on the Lengyel model [1 - 6]. A robust solution to power exhaust is one of the main challenges faced by reactor-scale fusion devices. To address the risk that a conventional divertor, as pursued in ITER, may not necessarily extrapolate to a DEMO reactor, the EUROfusion consortium has been systematically investigating alternative divertor configurations (ADCs) [1 – 5, 7, 8]. This effort has produced a large database of SOLPS-ITER simulations for the EU-DEMO with various divertor configurations [3 – 5, 9]. If accurate, the simple Lengyel model would provide an attractive approach to assess scrape-off layer (SOL) power exhaust and dissipation in scaling studies [10, 11]. However, many of the physical processes that are not included in the Lengyel model, such as cross-field transport, cooling due to interaction with the neutral population in the divertor, convective energy transport, or changes of flux expansion within the divertor leg, are the key features that separate ADCs from conventional configurations. Therefore, the Lengyel model is not expected to appropriately capture the key benefits of ADCs. In this study, it is observed that due to these missing physical processes, the Lengyel model overpredicts the threshold SOL impurity concentration for the onset of detachment in EU-DEMO by a factor of 5 – 10 relative to SOLPS-ITER simulations. Due to the dissipation of the peak heat flux by other processes than impurity radiation, such as cross-field transport to lower heat flux regions in the divertor, the total amount of power that must be dissipated by the impurity radiation is reduced by a factor of 2 – 4. Furthermore, one of the key assumptions of the Lengyel model is that heat is transported solely through electron heat conduction. Due to the strong temperature dependencies of the heat conduction and impurity cooling rates, the Lengyel model tends to predict a spatially very narrow radiative volume, limiting the total radiated power for a given impurity concentration. The SOLPS-ITER simulations indicate that the assumption of conduction dominated heat transport can be highly inaccurate in the region of divertor impurity radiation, such that the radiative volume and total dissipation for a given impurity concentration is enhanced compared to the pessimistic values predicted by the Lengyel model.
[1] F. Militello, et al. Nucl. Mat. Ene. 26 (2021) 100908.
[2] H. Reimerdes, et al. Nucl. Fusion 60 (2020) 066030.
[3] L. Xiang, et al. Nucl. Fusion 61 (2021) 076007.
[4] L. Aho-Mantila, et al. Nucl. Mat. Ene. 26 (2021) 100886.
[5] F. Subba, et al. Nucl. Mat. Ene. 12 (2017) 967-972.
[6] L. Lengyel, IPP Report 1/191, 1981.
[7] C. Theiler, et al. this conference.
[8] A. Thornton, et al. this conference.
[9] S. Wiesen, et al. J. Nucl. Mat. 463 (2015) 480-484.
[10] M.L. Reinke, et al. Nucl. Fusion 57 (2017) 034004.
[11] R.J. Goldston, et al. Plasma Phys. Control. Fusion 59 (2017) 055015.Speaker: Aaro Järvinen (VTT) -
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Detachment Control Considerations for Divertor Design
The control of a stable detachment solution that is compatible with both the core and edge may be crucial for the operation of reactor tokamaks in steady-state. In tokamaks there exist various methods of accessing and maintaining detachment, including fuelling, impurity seeding, and varying the heating power of the machine. How detachment is accessed and how its extent from the target (location of the detachment ‘front’) evolves with changes in those controllers can be significantly dependent on the characteristics of the configuration of the divertor employed.
In this work we introduce and utilize simple models that predict the location of a detachment front in a divertor, and how this location evolves with changes in controllers such as impurity seeding [Cowley C et al. Nuclear Fusion. 2022 Jun 20.] [Lipschultz B et al. Nuclear Fusion. 2016 Apr 8;56(5):056007.]. These simple models can be powerful tools to predict how changes in divertor configuration can affect access to and control over detachment. In this study, the simple model predictions have been compared to SOLPS-ITER simulations of isolated divertors in idealised geometries, including high power reactor-like simulations with artificial neon seeding. Broad agreement is found between the predictions of simple models and SOLPS-ITER, which may allow such models to be used for predicting detachment control in real machines.
The effects of features such as divertor flux expansion (total and poloidal), connection length, gradients in the total field along a flux tube, neutral baffling and magnetic pitch are analysed both in SOLPS-ITER simulation and simple modelling. An example of such analysis can be seen in the figures above, which show the movement of a detachment front in a straight grid and one which is poloidally flared near the target. The flared grid shows much slower poloidal front movement, and thus better detachment control in the flared region. However, in parallel space the movement of detachment fronts is similar in both grids, which implies that the difference in poloidal front movement is fundamentally due to connection length effects and magnetic pitch.
Along with this poloidally flared grid, many other divertor features are explored, and conclusions are drawn concerning how they can be used to influence the design of divertors for next-generation tokamak reactors and experiments.
Speaker: cyd Cowley (The University of York) -
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Impact of divertor geometry on separatrix density in JET H-mode plasmas and derivation of a scaling law as a function of engineering parameters
A viable magnetic fusion power plant has to combine very high plasma density and temperature in the core region, in order to maximize fusion reactions, with cold plasma conditions in the peripheral region compatible with long life expectancy of plasma-facing components. In this contribution, taking inspiration from recent work on DIIID tokamak (see ref. 1), we examine this crucial issue for magnetic fusion research by adopting an approach based on the analysis of a large set of experimental data on H-mode plasmas from JET tokamak. In order to obtain a scaling law for the relationship between top pedestal density and separatrix density at the outer midplane (OMP) as a function of engineering parameters more than ninety discharges have been considered. The choice to examine this density ratio is motivated by the fact that, on the one hand, the density at the top pedestal is an indicator of core confinement and device performance, and on the other hand, the separatrix density has a strong impact on divertor conditions, indicating whether safe conditions for divertor targets are achievable or not.
After a short description of the dataset under consideration and the power balance method used for the determination of the separatrix position [2], two main engineering parameters have been identified and used for the scaling law on the density ratio, namely the plasma current IP and the total injected power PTOTAL. This first scaling law seems to predict the experimental data quite well for low and medium values of the separatrix density, while at high density a strong discrepancy appears. In order to get further insight on such behavior the discharges were analyzed in terms of divertor magnetic configuration. A clear difference is observed between experiments with a corner-corner divertor configuration compared to the horizontal-vertical or vertical-vertical ones. This result suggests the introduction of a parameter taking into account the quality of confinement. In this way, a better agreement between predictions from the scaling law and experimental results is obtained for both low and high-density values [3]. Finally, numerical investigations for representative JET H-mode discharges in the three divertor configurations have been performed using the SOLEDGE code to analyze plasma conditions in the divertor region as well as in the main chamber and their impact on pedestal and separatrix density.Speaker: Guido CIRAOLO (CEA, IRFM) -
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SOLPS-ITER simulations of an X-point radiator in the single-null and snowflake divertor configurations in ASDEX Upgrade and EU-DEMO
The X-point radiator (XPR) is an attractive scenario to solve the power exhaust problem in future fusion devices. In ASDEX Upgrade (AUG), experiments with an XPR showed a dissipated power fraction larger than 90 %, fully detached divertor targets and ELM suppression with a moderate confinement degradation [1]. Recently, a reduced model [2] was derived to explain the physical mechanisms for initiating a stable XPR, highlighting the role of neutrals at the X-point and the magnetic connection length and flux expansion.
In this work, the 2D transport code SOLPS-ITER [3] was applied to reproduce the XPR phenomenon. The simulation results show qualitative agreements with the experimental measurements and the reduced model in AUG. By analyzing the particle, momentum and energy balances, the particle and heat transport in an XPR are depicted in detail. In addition to this, the important role of neutrals and the magnetic connection length in the generation of an XPR are demonstrated by the simulations with virtual neutral baffles and with various toroidal magnetic field strength, respectively.
An XPR regime was also achieved in SOLPS-ITER simulations of the snowflake divertor configurations in the new upper divertor of AUG [4] as well as in the alternative divertor concept of EU-DEMO [5]. With the substantially larger connection length and flux expansion compared to the conventional single-null divertor, the snowflake divertor showed an easier access of the XPR regime with a lower impurity concentration at the separatrix.
In order to demonstrate the feasibility of the XPR regime in future devices, a machine-size scaling of the access condition, the impurity compression and the maximum radiative fraction in the X-point volume is highly required. This remains an open question and will be discussed in this presentation.
References
[1] M. Bernert, et al., Nucl. Fusion 61, 024001 (2021)
[2] U. Stroth, et al., Nucl. Fusion 62, 076008 (2022)
[3] S. Wiesen, et al., J. Nucl. Mater. 463, 480 (2015)
[4] T. Lunt, et al., Nucl. Mater. Energy 12, 1037 (2017)
[5] H. Reimerdes, et al., Nucl. Fusion 60, 066030 (2020)Speaker: Ou Pan (Max-Planck-Institut für Plasmaphysik) -
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Progress towards robust divertor and exhaust scenario optimization with SOLPS-ITER
Divertor and exhaust scenario design for future reactors such as ITER and DEMO heavily rely on plasma edge codes as SOLPS-ITER [1]. In practice, the design process often proceeds through large parameter scans to explore the operational space, whereby divertor shape, magnetic field and model parameters are manually tuned to improve the performance of the design and meet various physics, engineering and material constraints. Combined with the complex and computationally expensive nature of plasma edge codes, this procedure is extremely cumbersome. Moreover, uncertainty on model and design parameters is often not accounted for explicitly.
Gradient-based optimization methods can significantly reduce the design effort, by automatically finding the design that optimizes performance, which is quantified by a cost function. Using adjoint techniques, the required design sensitivities are computed at a cost independent of the number of design variables. In the past years it has been demonstrated that these methods can be used to find optimal target shapes and magnetic fields minimizing the peak power loads towards the divertor based on somewhat simplified edge models [2]. In this contribution, we discuss recent progress towards enabling these methods for complex edge codes.
For efficient and accurate sensitivity computations in complex, continuously advancing codes, we use Algorithmic Differentiation (AD) which provides semiautomatic derivatives by direct processing of the source code. Both forward and adjoint modes of AD are applied successfully to SOLPS-ITER [3] using the TAPENADE tool [4]. Using these sensitivity tools, a wealth of information becomes accessible to the designer, allowing to identify the input parameters with largest impact on the solution, and providing full 2D sensitivity maps showing the spatial dependence. We use the same AD-based optimization algorithms in a Bayesian context to calibrate the model parameters based on experimental data [5], providing also information on the uncertainty distribution of the parameters. To enable divertor shape optimization in SOLPS-ITER, a grid smoothing tool is being developed to provide the necessary grids during the optimization. These achievements further open the way to robust design procedures, which optimize performance taking uncertainty into account.
A second challenge relates to accurate sensitivity evaluations in the presence of Monte Carlo noise from the kinetic neutral simulation. Without special measures, the sensitivities would be overwhelmed by the statistical noise. By differentiating the individual particle trajectories at fixed random numbers, however, reliable sensitivities are found [6]. AD has great potential in this context.
These optimization schemes are being integrated in the recently developed extended grids version of SOLPS-ITER, which also contains various enhanced model options to improve the accuracy and speed up individual edge simulations, optimization, and uncertainty quantification at the reactor scale.
[1] A.S. Kukushkin et al., Fusion Eng. Des. 86 (2011) 2865
[2] M. Baelmans et al., Nucl. Fusion 57 (2017) 036022
[3] S. Carli et al., Nucl. Mater. Energy 18 (2018) 6
[4] L. Hascoet et al., ACM Trans. Math. Softw. 39 (2013) 20
[5] S. Carli et al., Contrib. Plasma Phys. (2021) e202100184
[6] W. Dekeyser et al., Contrib. Plasma Phys. 58 (2018) 643Speaker: Wouter Dekeyser (KU Leuven, Department of Mechanical Engineering) -
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EIRENE modelling with improved CRMs for spectroscopic RT detachment control in EU-DEMO
EU-DEMO reactor operation is expected to go beyond the ITER requirements of a semi-detached divertor regime [1], and will have to achieve the even higher levels of dissipation needed to demonstrate a realistic power exhaust solution for a fusion power plant. Sustaining the desired degree of detachment will require reliable real time (RT) control. In addition to the actuators (e.g. the impurity seeding rate) one needs reliable, but robust in use (fast, monotonic, free of additional dependences etc) diagnostic signal(s) to characterise the detachment state. Line-of-sight (LOS) integrated spectroscopy is a promising way to provide such a signal, however a deep physics understanding and mature modelling basis is needed for interpretation of the LOS integrated intensities as well as for suggesting the most optimal spectroscopic features (e.g. particular line ratios) to be employed. This work is focused on the related refinement of established modelling tools required for such purposes, namely on development of the atomic and molecular (A&M) collisional-radiative models (CRM) for the EIRENE neutral Monte-Carlo solver, which can also be utilized standalone for investigating A&M effects on a fixed background plasma.
The impact of molecular release of particles from the wall and further reaction sequences in the plasma are known to play a key role in divertor operations including the onset and parameters of detachment. Rotational and vibrational temperatures influence the spectroscopic interpretation of LOS signals. A new CRM [2] is demonstrated that captures part of those effects and provides the regime-changing effects on branching ratios for hydrogen specie emission as a function of plasma parameters. The “free” parameters of the model (H2 source) are calibrated with respect to the underlying SOLPS-ITER (B2.5-EIRENE) EU-DEMO edge plasma backgrounds [3].
Another challenge for simulations on DEMO scale is the overall code performance, which can be improved through new physics features like fluid-kinetic hybridisation of the neutral model, as well as new parallelization paradigms in EIRENE as demonstrated in Ref. [4]. The use of the CRM inside the code can also bear on performance, and this can be partially addressed through following up species internal state variables (sparing the necessity to follow the resolved states as separate species, but implying an approximation in the description of the transport).
This work is a part of the EUROfusion E-TASC activity [5] and feeds into the EU-DEMO exhaust modelling roadmap [6]. Those include interfaces to relevant codes, unification of the underlying data, IMASification etc. Part of the project is focussing also on visualisation tools and synthetic diagnostics including the LOS in the context of 3D DEMO divertor geometry assessment. The modelling can assist in future the EU-DEMO diagnostic design by selecting the optimal number and geometry of the LOS by means of demonstrating reliable synthetic detachment control schemes and predictions.
[1] R.A.Pitts et al., 2019 NME, vol.20, 100696
[2] F.Cianfrani et al., EPS-2022, P2a.105
[3] F.Subba et al. 2020 NF 61 106013
[4] D.V.Borodin et al., 2022 NF 62 086051
[5] X Litaudon et al 2022 PPCF 64 034005
[6] S.Wiesen et al., this conferenceSpeaker: Dr Dmitriy V. Borodin (Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), 52425 Jülich, Germany)
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Discussion Session: Modelling Board Room A
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Convener: Bruce Lipschultz (University of York) -
Closed session [PC Meeting] MOE05
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Dinner [tbc] Board Room A
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Radiative Divertor Board Room A
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Convener: Matthew Reinke (Commonwealth Fusion Systems)-
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Entry to and exit from ELM suppressed H-mode in detached conditions
While detachment is in present-day machines routinely achieved under different conditions during the flattop of a discharge by either density ramps or extrinsic impurity seeding, it has not yet been shown to be accessed and controlled throughout the full duration of a discharge. For future machines, it is essential to be in detachment already during the power ramp up, the L-H transition and also ramp down, in order to assure that the heat load limit of the target materials are not exceeded.
During these transient conditions, the heat flux to the target is continuously changing, e.g due to changes of the exhaust power, changes of the power flux width or due to movements of the strikeline. This makes the control of detachment and the guaranteed safe divertor operation challenging.
At ASDEX Upgrade, detachment can be controlled either by proxies of the electron temperature at the target (partial detachment) or by the location of a so-called X-point radiator XPR (full detachment). For the latter, it was shown that power transients can be well buffered. In combination with well-adjusted pre-seeding, allowing to detect the XPR in L-mode, it is possible to achieve detachment before the L-H transition. With the activated XPR controller, the height of XPR relative to the X-point is regulated, and, thus, detachment maintained during the increase of the injected power, the resulting L-H transition and also throughout the flattop phase of the discharge. The controller also assures to maintain detachment at a safe level during the power and current ramp down. In this contribution we will present the recent demonstration of these capabilities within one discharge at ASDEX Upgrade and discuss the reproducibility of detached L-H transitions in different scenarios.
The experiments show for the first time that an H-mode can be accessed while avoiding too high heat loads on the target beforehand. Further input from the community is required to expand this proof of principle to meet all requirements for the current ramp-up and -down in a future reactor.
Speaker: Dr Matthias Bernert (Max-Planck-Institut für Plasmaphysik) -
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Particle and power exhaust of new EAST lower tungsten divertor for advanced steady-state operations
Significant progress has been made on the new lower tungsten (W) divertor with closed geometry and active water-cooling capability for steady state operations in EAST since 2021. The latest experimental results demonstrate that the new divertor exhibits strong particle exhaust capability and relatively high neutral retention in the divertor region, which facilitate both impurity screening and divertor detachment with respect to the less closed upper divertor. By using the new W divertor, the divertor detachment with a strong reduction of particle and heat fluxes has been achieved in lower single null (LSN) H-mode plasmas with the ion B×▽B drift towards the X-point. Compared with the upper divertor, the more closed lower divertor has a lower detachment density threshold. When the strike point locates on the horizontal target with divertor closure increasing, the detachment can be accessed more easily, which are in good agreement with the simulations during divertor design.
The impurity seeding with both argon (Ar) and neon (Ne) to reduce heat load has been performed in high-performance plasmas by leveraging the effect of drifts and impurity seeding location, and optimized divertor configuration coupled with strong pumping. A confinement improvement was observed with Ne seeding. For Ar seeding, the result indicates that Ar is more efficient at cooling electron temperature on divertor targets, which can lead to simultaneous enhancement of core and divertor radiation, accompanied with confinement degradation. In addition, the integration of large-ELM-suppression and detachment with H_98 ~ 1 has been achieved with the new divertor configuration and Ne seeding. For active long pulse detachment control, 30s H-mode operation with a detachment-control duration of 25s has been achieved in EAST. A series of detachment or radiation feedback control techniques for core-edge integration have been further developed and demonstrated in long-pulse H-mode plasmas. During the feedback control phase, the plasma stored energy was well maintained at a stable level with T_et ~ 5 eV near the strike point and H_98 > 1. These experiments demonstrated good compatibility of high core plasma performance with divertor detachment. It thus offers a highly promising plasma control scenario suitable for long-pulse high-performance H-mode operation in EAST, which is potentially applicable to future fusion reactors.[1] L. Wang et al., Nucl. Fusion 62, 076002 (2022)
[2] G. S. Xu et al., Nucl. Fusion 61, 126070 (2021)
[3] L. Y. Meng et al., Nucl. Fusion 62, 086027 (2022)Speaker: Dr Liang Wang (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP)) -
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Radiative divertor compatible with RMP-driven, ELM-crash-suppression in fusion DEMO-type devices
The edge-localized-modes (ELMs) in magnetic fusion reactor raise a major concern not only due to the degraded core confinement, but also due to the adverse effects against plasma facing components [1]. In the existing medium size tokamaks, the application of resonant magnetic perturbation (RMP) has proven effective in suppressing ELMs without being limited to ELM-crash-mitigation. Nonetheless, the RMP-driven, ELM-crash-suppression often elevates the peak of divertor heat flux at least by a factor of 2 or 3, in comparison with that of inter-ELMs without RMPs [2]. In that regard, the impurity seeding and deuterium gas puffing in the periphery of the plasma column had been expected to play a role in releasing the excessive energy in a radiative manner, which is essential to reach detached plasmas in divertor power exhaust handling.
However, whenever RMP-driven, ELM-crash-suppression becomes gradually radiative with either impurity seeding or gas puffing, it loses the high quality of ELM-crash-suppression. On the other hand, once high density, near-detached plasmas are accessed by RMPs, it has been nearly impossible, if not forbidden, to achieve the RMP-driven, ELM-crash-suppression at least based on the KSTAR experiments for more than a decade. Interestingly, in terms of ‘wetted area’, the strongly mitigated ELM-crashes in radiative divertor could be more favorable than the full ELM-crash-suppression [3]. For that reason, it is quite important to address whether radiative divertor, which is expected to be a routine in fusion reactor, would be compatible with RMP-driven, ELM-crash-suppression sooner rather than later. In recent KSTAR experiments, a series of promising results show that ITER-like 3-row RMP-driven, ELM-crash-suppression would be much better than 2-row counterparts in terms of the divertor thermal loading, while such 3-D configuration would be more amenable to radiative divertor [4]. At the same time, depending on the scrape-off-layer conditions, both RMP-driven, ELM-crash-suppression and detachment conditions vary significantly. Considering a preliminary K-DEMO design study shows an enormous challenge of power handling capability in divertor [5], a more comprehensive understanding of radiative divertor should be prioritized in reactor study, while its compatibility with ex-vessel RMP control needs to be addressed, if standard H-mode with ELMs remains a baseline scenario in fusion reactor.References
[1] A. Loarte et al, Nucl. Fusion 54 033007 (2014)
[2] Y. In et al, Nucl. Fusion 59, 056009 (2019)
[3] Y. In et al, Nucl. Fusion 62, 066014 (2022)
[4] Y. In et al, Plenary talk at AAPPS-DPP (2021)
[5] S.J. Kwon et al, Fusion Engineering and Design 159, 111770 (2020)Speaker: Prof. Yongkyoon In (Ulsan National Institute of Science and Technology)
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Radiative Divertor Board Room A
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Convener: Matthew Reinke (Commonwealth Fusion Systems)-
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Divertor detachment and re-attachment studies with mixed impurity seeding on ASDEX Upgrade
In GW-scale fusion power plants, most of the alpha heating power must be radiated from the edge plasma on closed field lines and the remaining fraction radiated in the scrape-off layer to reduce the surface heat flux and material erosion rate in the divertor below tolerable values. A mixture of impurities with radiation efficiencies tailored to the distinctly different regions of plasma within the tokamak will likely be injected to achieve this level of radiation. Here we overview a set of experiments on ASDEX Upgrade (AUG) that focus on divertor detachment in ELMy H-mode plasmas with different mixtures of seeded N, Ne, and Ar gas. The scenarios have a plasma current of 1 MA, powers crossing the separatrix up to 10 MW, and divertor neutral pressures ranging up to ~4 Pa. The experiments explore both the transient divertor response to power modulations and temporary impurity gas cuts and the steady state conditions required for partial and pronounced detachment with mixtures of N and Ne and N and Ar, including comparison to equivalent scenarios with single impurity injection. Contrary to previous experience on AUG, a stable detached scenario with good core plasma performance was achieved with Ne seeding when combined with a low level of N seeding.
To understand the physics that determines the timescale for divertor re-attachment, power perturbations of ~4 MW introduced by additional NBI heating were applied during steady-state phases with different degrees of divertor detachment. In phases of partial detachment and detachment onset the divertor re-attached within ~80 ms, similar to the energy confinement time. However, the divertor took ~250 ms to re-attach in the phase with pronounced detachment and strong radiation near the X-point. A simple model to predict the re-attachment time has been developed considering timescales of atomic processes and geometry. Similar experiments were run replacing the power modulation with cuts in impurity seeding gas. In these experiments, the timescale for re-attachment appears to be set by the impurity residence time; however, work is ongoing to disentangle the different physics mechanisms at play. These results should open discussion regarding the minimum timescales that real-time control systems in future reactors must be able to actuate on to react to transients.
In the steady-state phases, an analytical formula for the impurity seeded partially detached divertor operational point developed by Kallenbach et al. is extended to include impurity mixtures and compared to experimental values derived from divertor spectroscopy and neutral pressure measurements. Focus is given to the radiation efficiency fractions used for each impurity, which were previously only estimated for Ne and Ar. The formula gives satisfactory agreement to the measurements across the range of impurity mixtures and divertor conditions. While the formula is valid only for partial detachment, threshold conditions for pronounced detached are again found not far beyond the partial detachment threshold limit. Understanding how to account for impurity mixtures provides an opportunity for combining impurity concentration measurements from both W- and C- wall machines to assess the machine size scaling of the detachment threshold.
Speaker: Stuart Henderson (UKAEA) -
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Overview of Exhaust Physics Results from MAST-U
The particle and power exhaust solution in fusion reactors needs to produce tolerable power loads on plasma-facing-components and be compatible with the high-performance fusion core. Alternative divertor configurations should be explored to mitigate the risk that the ITER solution, a single null divertor with a high radiation fraction in the scrape-off-layer (SOL), does not extrapolate to future higher power devices. MAST-U was designed to explore the comparative benefits of conventional and alternative configurations, in particular the Super-X configuration [1] with tightly baffled divertor chambers. The Super-X configuration increases the outer strike point major radius by a factor of ~2, which increases the connection length and total flux expansion by roughly the same factor. This reduces the heat flux to the outer target and improves access to the detached divertor regime due to the combined effect of increased wetted area and enhanced volumetric power losses and cross-field transport.
During the first MAST-U campaign, the aim of the main exhaust experiments was to demonstrate the expected advantages of the Super-X configuration compared with a conventional divertor. An ohmic heated L-mode scenario in double-null configuration with plasma current of 600-750 kA and elongation of ~2 was developed for both conventional and Super-X configurations. Midplane electron temperature and electron density profiles were similar for the two configurations, which shows that the core plasma conditions were not adversely affected by the Super-X configuration or the presence of detached divertors. The ion flux rollover at the outer target measured by Langmuir probes was 50% lower for the Super-X configuration in terms of midplane line average density, in agreement with previous SOLPS and analytic modelling predictions [2]. During attached divertor conditions, the peak heat flux measured by IR thermography was a factor of ~10 lower for the Super-X configuration. Multi-wavelength imaging and spectroscopy diagnostics identified distinct regions of ionisation, molecular activated recombination and electron-ion recombination in the divertor chamber during the detachment evolution. Results from the SOLPS-ITER code indicate (i) particle drifts do not seem to play a major role in the outer scrape-off-layer; (ii) the outer target ion flux rollover in the conventional configuration is driven by power loss from the core, rather than from the divertor; (iii) unlike our expectations for a reactor, the core and main SOL particle sources are significant compared to the target flux. This motivates going to higher heating powers in future experiments.
The aim of this contribution is to give an overview of exhaust physics results from MAST-U and discuss the implications for future reactors (enhanced role of plasma-molecular interactions, simulation-experiment comparison in current devices is essential for credible extrapolation to more reactor relevant conditions). Results from the first campaign will be presented along with preliminary analysis from the second campaign, starting in September 2022. The main exhaust objective for the second campaign is to study the compatibility of high-confinement core scenarios with optimised conventional and Super-X configurations.
[1] P. M. Valanju et al Phys. Plasmas 16 056110 (2009)
[2] B. Lipschultz et al Nucl. Fusion 56 056007 (2016)Speaker: Dr Peter Ryan (UKAEA) -
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Compact radiative divertor experiments at ASDEX Upgrade extrapolated to DEMO
One of the most promising approaches to tackle the power exhaust problem in a divertor tokamak is the so called X-point radiator (XPR). By the controlled injection of impurities into the plasma a radiation cloud localized in the vicinity of the X-point is formed. It was shown that up to 95% of the power absorbed in the plasma can be dissipated before reaching the divertor targets [1]. Recent experiments [2], an analytic model [3] and numerical simulations with SOLPS-ITER [4] have shown that the temperature at the X-point under these conditions can become as low as a few eV,similar to the conditions in a detached divertor. We here report on experiments that demonstrate,that with an XPR long divertor legs and complicated divertor geometries may not be needed anymore. After establishing an XPR via nitrogen seeding in an H-mode with up to 15 MW of total heating power the X-point was moved towards the tungsten target surface such that the divertor legs become as short as 5 cm and the poloidal flux expansion as large as $f_x=ds_t/dr_u=50$. We also refer to this configuration as a `compact radiative divertor' (CRD). The energy confinement remained constant during the movement. The challenge of this configurations is that the (projected) field line incidence angle $\theta_\perp$ is inversely proportional to $f_x$ and therefore $\theta_\perp$ very small. Commonly it is assumed in the community that $\theta_\perp$ must be kept above $\sim2^\circ$ otherwise magnetic error fields or small misalignments of the divertor tiles could lead to strong toroidal asymmetries in the power deposition pattern or even to the formation of hot spots. At least with respect to the error fields it has recently been found in 3D transport simulations that the shallow angles might not be as problematic as assumed [5] when detachment can be guaranteed. The experiments presented here now confirm this finding: Despite of values for $\theta_\perp$ in the order of $0.2^\circ$ in the CRD configuration no hot spots were observed in the near-SOL region observed by the camera. Even with the X-point located exactly on the target surface and even without density or impurity feed-back control the discharge remained stable, hot spots absent and the divertor in a detached state. The maximized volume of the confined plasma, the smaller poloidal field coil currents and the increased vertical stability of the CRD are all very attractive properties for DEMO. In DEMO the XPR volume is of far higher relevance for power exhaust due to the enhanced ratio $a/\lambda_q$, compared to AUG. Further research will be devoted to the question on how to realize a pumping concept for He-ash removal and on how to access the high power CRD without passing a phase with harmful divertor conditions. At least at very high densities, an L-H transition with a detached CRD was observed indicating a possible path to this configuration without attaching the divertor.
[1] Bernert, NF 2020
[2] Cavedon, NF 2022
[3] Stroth, NF 2022
[4] Pan, to be submitted to NF 2022
[5] Lunt, Nucl.Mat.Energ. 2021Speaker: Tilmann Lunt (Max-Planck-Institut für Plasmaphysik)
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Convener: Matthew Reinke (Commonwealth Fusion Systems) -
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Convener: Marco Wischmeier (IPP Garching) -
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Conveners: Marco Wischmeier (IPP Garching), Matteo Barbarino (International Atomic Energy Agency)
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